Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 22563-22576 [2013-08756]
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Federal Register / Vol. 78, No. 73 / Tuesday, April 16, 2013 / Notices
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Submit, by June 17, 2013, comments
that address the following questions:
1. Is the proposed collection of
information necessary for the NRC to
properly perform its functions? Does the
information have practical utility?
2. Is the burden estimate accurate?
3. Is there a way to enhance the
quality, utility, and clarity of the
information to be collected?
4. How can the burden of the
information collection be minimized,
including the use of automated
collection techniques or other forms of
information technology?
The public may examine and have
copied for a fee publicly available
documents, including the draft
supporting statement, at the NRC’s
Public Document Room, Room O–1F21,
One White Flint North, 11555 Rockville
Pike, Rockville, Maryland 20852. The
OMB clearance requests are available at
the NRC’s Web site: https://www.nrc.gov/
public-involve/doc-comment/omb/.
The document will be available on the
NRC’s home page site for 60 days after
the signature date of this notice.
Comments submitted in writing or in
electronic form will be made available
for public inspection. Because your
comments will not be edited to remove
any identifying or contact information,
the NRC cautions you against including
any information in your submission that
you do not want to be publicly
disclosed. Comments submitted should
reference Docket No. NRC–2013–0054.
You may submit your comments by
any of the following methods: Electronic
comments: Go to https://
www.regulations.gov and search for
Docket No. NRC–2013–0054. Mail
comments to NRC Clearance Officer,
Tremaine Donnell (T–5 F53), U.S.
Nuclear Regulatory Commission,
Washington, DC 20555–0001. Questions
about the information collection
requirements may be directed to the
NRC Clearance Officer, Tremaine
Donnell (T–5 F53), U.S. Nuclear
Regulatory Commission, Washington,
DC 20555–0001, by telephone at 301–
415–6258, or by email to
INFOCOLLECTS.Resource@NRC.GOV.
I. Accessing Information and
Submitting Comments
NUCLEAR REGULATORY
COMMISSION
A. Accessing Information
[NRC–2013–0069]
Biweekly Notice; Applications and
Amendments to Facility Operating
Licenses and Combined Licenses
Involving No Significant Hazards
Considerations
Background
Pursuant to Section 189a. (2) of the
Atomic Energy Act of 1954, as amended
(the Act), the U.S. Nuclear Regulatory
Commission (NRC) is publishing this
regular biweekly notice. The Act
requires the Commission publish notice
of any amendments issued, or proposed
to be issued and grants the Commission
the authority to issue and make
immediately effective any amendment
to an operating license or combined
license, as applicable, upon a
determination by the Commission that
such amendment involves no significant
hazards consideration, notwithstanding
the pendency before the Commission of
a request for a hearing from any person.
This biweekly notice includes all
notices of amendments issued, or
proposed to be issued from March 21 to
April 3, 2013. The last biweekly notice
was published on April 2, 2013 (78 FR
19746).
[FR Doc. 2013–08875 Filed 4–15–13; 8:45 am]
You may access information
and comment submissions related to
this document, which the NRC
possesses and is publicly-available, by
searching on https://www.regulations.gov
under Docket ID NRC–2013–0069. You
may submit comments by any of the
following methods:
• Federal Rulemaking Web site: Go to
https://www.regulations.gov and search
for Docket NRC–2013–0069. Address
questions about NRC dockets to Carol
Gallagher; telephone: 301–492–3668;
email: Carol.Gallagher@nrc.gov.
• Mail comments to: Cindy Bladey,
Chief, Rules, Announcements, and
Directives Branch (RADB), Office of
Administration, Mail Stop: TWB–05–
B01M, U.S. Nuclear Regulatory
Commission, Washington, DC 20555–
0001.
• Fax comments to: RADB at 301–
492–3446.
For additional direction on accessing
information and submitting comments,
see ‘‘Accessing Information and
Submitting Comments’’ in the
SUPPLEMENTARY INFORMATION section of
this document.
BILLING CODE 7590–01–P
SUPPLEMENTARY INFORMATION:
Dated at Rockville, Maryland, this 10th day
of April 2013.
For the Nuclear Regulatory Commission.
Tremaine Donnell,
NRC Clearance Officer, Office of Information
Services.
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ADDRESSES:
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Please refer to Docket ID NRC–2013–
0069 when contacting the NRC about
the availability of information regarding
this document. You may access
information related to this document,
which the NRC possesses and is
publicly available, by the following
methods:
• Federal Rulemaking Web site: Go to
https://www.regulations.gov and search
for Docket ID NRC–2013–0069.
• NRC’s Agencywide Documents
Access and Management System
(ADAMS): You may access publicly
available documents online in the NRC
Library at https://www.nrc.gov/readingrm/adams.html. To begin the search,
select ‘‘ADAMS Public Documents’’ and
then select ‘‘Begin Web-based ADAMS
Search.’’ For problems with ADAMS,
please contact the NRC’s Public
Document Room (PDR) reference staff at
1–800–397–4209, 301–415–4737, or by
email to pdr.resource@nrc.gov.
Documents may be viewed in ADAMS
by performing a search on the document
date and docket number.
• NRC’s PDR: You may examine and
purchase copies of public documents at
the NRC’s PDR, Room O1–F21, One
White Flint North, 11555 Rockville
Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC–2013–
0069 in the subject line of your
comment submission, in order to ensure
that the NRC is able to make your
comment submission available to the
public in this docket.
The NRC cautions you not to include
identifying or contact information in
comment submissions that you do not
want to be publicly disclosed. The NRC
posts all comment submissions at https://
www.regulations.gov as well as entering
the comment submissions into ADAMS,
and the NRC does not edit comment
submissions to remove identifying or
contact information.
If you are requesting or aggregating
comments from other persons for
submission to the NRC, then you should
inform those persons not to include
identifying or contact information in
their comment submissions that they do
not want to be publicly disclosed. Your
request should state that the NRC will
not edit comment submissions to
remove such information before making
the comment submissions available to
the public or entering the comment
submissions into ADAMS.
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Federal Register / Vol. 78, No. 73 / Tuesday, April 16, 2013 / Notices
Notice of Consideration of Issuance of
Amendments to Facility Operating
Licenses and Combined Licenses,
Proposed No Significant Hazards
Consideration Determination, and
Opportunity for a Hearing
The Commission has made a
proposed determination that the
following amendment requests involve
no significant hazards consideration.
Under the Commission’s regulations in
§ 50.92 of Title 10 of the Code of Federal
Regulations (10 CFR), this means that
operation of the facility in accordance
with the proposed amendment would
not (1) involve a significant increase in
the probability or consequences of an
accident previously evaluated; or (2)
create the possibility of a new or
different kind of accident from any
accident previously evaluated; or (3)
involve a significant reduction in a
margin of safety. The basis for this
proposed determination for each
amendment request is shown below.
The Commission is seeking public
comments on this proposed
determination. Any comments received
within 30 days after the date of
publication of this notice will be
considered in making any final
determination.
Normally, the Commission will not
issue the amendment until the
expiration of 60 days after the date of
publication of this notice. The
Commission may issue the license
amendment before expiration of the 60day period provided that its final
determination is that the amendment
involves no significant hazards
consideration. In addition, the
Commission may issue the amendment
prior to the expiration of the 30-day
comment period should circumstances
change during the 30-day comment
period such that failure to act in a
timely way would result, for example in
derating or shutdown of the facility.
Should the Commission take action
prior to the expiration of either the
comment period or the notice period, it
will publish in the Federal Register a
notice of issuance. Should the
Commission make a final No Significant
Hazards Consideration Determination,
any hearing will take place after
issuance. The Commission expects that
the need to take this action will occur
very infrequently.
Within 60 days after the date of
publication of this notice, any person(s)
whose interest may be affected by this
action may file a request for a hearing
and a petition to intervene with respect
to issuance of the amendment to the
subject facility operating license or
combined license. Requests for a
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hearing and a petition for leave to
intervene shall be filed in accordance
with the Commission’s ‘‘Agency Rules
of Practice and Procedure’’ in 10 CFR
Part 2. Interested person(s) should
consult a current copy of 10 CFR 2.309,
which is available at the NRC’s PDR,
located at One White Flint North, Room
O1–F21, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852. The
NRC regulations are accessible
electronically from the NRC Library on
the NRC’s Web site at https://
www.nrc.gov/reading-rm/doccollections/cfr/. If a request for a hearing
or petition for leave to intervene is filed
by the above date, the Commission or a
presiding officer designated by the
Commission or by the Chief
Administrative Judge of the Atomic
Safety and Licensing Board Panel, will
rule on the request and/or petition; and
the Secretary or the Chief
Administrative Judge of the Atomic
Safety and Licensing Board will issue a
notice of a hearing or an appropriate
order.
As required by 10 CFR 2.309, a
petition for leave to intervene shall set
forth with particularity the interest of
the petitioner in the proceeding, and
how that interest may be affected by the
results of the proceeding. The petition
should specifically explain the reasons
why intervention should be permitted
with particular reference to the
following general requirements: (1) The
name, address, and telephone number of
the requestor or petitioner; (2) the
nature of the requestor’s/petitioner’s
right under the Act to be made a party
to the proceeding; (3) the nature and
extent of the requestor’s/petitioner’s
property, financial, or other interest in
the proceeding; and (4) the possible
effect of any decision or order which
may be entered in the proceeding on the
requestor’s/petitioner’s interest. The
petition must also identify the specific
contentions which the requestor/
petitioner seeks to have litigated at the
proceeding.
Each contention must consist of a
specific statement of the issue of law or
fact to be raised or controverted. In
addition, the requestor/petitioner shall
provide a brief explanation of the bases
for the contention and a concise
statement of the alleged facts or expert
opinion which support the contention
and on which the requestor/petitioner
intends to rely in proving the contention
at the hearing. The requestor/petitioner
must also provide references to those
specific sources and documents of
which the petitioner is aware and on
which the requestor/petitioner intends
to rely to establish those facts or expert
opinion. The petition must include
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sufficient information to show that a
genuine dispute exists with the
applicant on a material issue of law or
fact. Contentions shall be limited to
matters within the scope of the
amendment under consideration. The
contention must be one which, if
proven, would entitle the requestor/
petitioner to relief. A requestor/
petitioner who fails to satisfy these
requirements with respect to at least one
contention will not be permitted to
participate as a party.
Those permitted to intervene become
parties to the proceeding, subject to any
limitations in the order granting leave to
intervene, and have the opportunity to
participate fully in the conduct of the
hearing.
If a hearing is requested, the
Commission will make a final
determination on the issue of no
significant hazards consideration. The
final determination will serve to decide
when the hearing is held. If the final
determination is that the amendment
request involves no significant hazards
consideration, the Commission may
issue the amendment and make it
immediately effective, notwithstanding
the request for a hearing. Any hearing
held would take place after issuance of
the amendment. If the final
determination is that the amendment
request involves a significant hazards
consideration, then any hearing held
would take place before the issuance of
any amendment.
All documents filed in NRC
adjudicatory proceedings, including a
request for hearing, a petition for leave
to intervene, any motion or other
document filed in the proceeding prior
to the submission of a request for
hearing or petition to intervene, and
documents filed by interested
governmental entities participating
under 10 CFR 2.315(c), must be filed in
accordance with the NRC’s E-Filing rule
(72 FR 49139; August 28, 2007). The EFiling process requires participants to
submit and serve all adjudicatory
documents over the internet, or in some
cases to mail copies on electronic
storage media. Participants may not
submit paper copies of their filings
unless they seek an exemption in
accordance with the procedures
described below.
To comply with the procedural
requirements of E-Filing, at least 10
days prior to the filing deadline, the
participant should contact the Office of
the Secretary by email at
hearing.docket@nrc.gov, or by telephone
at 301–415–1677, to request (1) a digital
identification (ID) certificate, which
allows the participant (or its counsel or
representative) to digitally sign
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documents and access the E-Submittal
server for any proceeding in which it is
participating; and (2) advise the
Secretary that the participant will be
submitting a request or petition for
hearing (even in instances in which the
participant, or its counsel or
representative, already holds an NRCissued digital ID certificate). Based upon
this information, the Secretary will
establish an electronic docket for the
hearing in this proceeding if the
Secretary has not already established an
electronic docket.
Information about applying for a
digital ID certificate is available on the
NRC’s public Web site at https://
www.nrc.gov/site-help/e-submittals/
apply-certificates.html. System
requirements for accessing the ESubmittal server are detailed in the
NRC’s ‘‘Guidance for Electronic
Submission,’’ which is available on the
agency’s public Web site at https://
www.nrc.gov/site-help/esubmittals.html. Participants may
attempt to use other software not listed
on the Web site, but should note that the
NRC’s E-Filing system does not support
unlisted software, and the NRC Meta
System Help Desk will not be able to
offer assistance in using unlisted
software.
If a participant is electronically
submitting a document to the NRC in
accordance with the E-Filing rule, the
participant must file the document
using the NRC’s online, Web-based
submission form. In order to serve
documents through the Electronic
Information Exchange System, users
will be required to install a Web
browser plug-in from the NRC’s Web
site. Further information on the Webbased submission form, including the
installation of the Web browser plug-in,
is available on the NRC’s public Web
site at https://www.nrc.gov/site-help/esubmittals.html.
Once a participant has obtained a
digital ID certificate and a docket has
been created, the participant can then
submit a request for hearing or petition
for leave to intervene. Submissions
should be in Portable Document Format
(PDF) in accordance with the NRC
guidance available on the NRC’s public
Web site at https://www.nrc.gov/sitehelp/e-submittals.html. A filing is
considered complete at the time the
documents are submitted through the
NRC’s E-Filing system. To be timely, an
electronic filing must be submitted to
the E-Filing system no later than 11:59
p.m. Eastern Time on the due date.
Upon receipt of a transmission, the EFiling system time-stamps the document
and sends the submitter an email notice
confirming receipt of the document. The
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E-Filing system also distributes an email
notice that provides access to the
document to the NRC’s Office of the
General Counsel and any others who
have advised the Office of the Secretary
that they wish to participate in the
proceeding, so that the filer need not
serve the documents on those
participants separately. Therefore,
applicants and other participants (or
their counsel or representative) must
apply for and receive a digital ID
certificate before a hearing request/
petition to intervene is filed so that they
can obtain access to the document via
the E-Filing system.
A person filing electronically using
the agency’s adjudicatory E-Filing
system may seek assistance by
contacting the NRC Meta System Help
Desk through the ‘‘Contact Us’’ link
located on the NRC’s Web site at https://
www.nrc.gov/site-help/esubmittals.html, by email at
MSHD.Resource@nrc.gov, or by a tollfree call at 1–866 672–7640. The NRC
Meta System Help Desk is available
between 8 a.m. and 8 p.m., Eastern
Time, Monday through Friday,
excluding government holidays.
Participants who believe that they
have a good cause for not submitting
documents electronically must file an
exemption request, in accordance with
10 CFR 2.302(g), with their initial paper
filing requesting authorization to
continue to submit documents in paper
format. Such filings must be submitted
by: (1) First class mail addressed to the
Office of the Secretary of the
Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555–
0001, Attention: Rulemaking and
Adjudications Staff; or (2) courier,
express mail, or expedited delivery
service to the Office of the Secretary,
Sixteenth Floor, One White Flint North,
11555 Rockville Pike, Rockville,
Maryland, 20852, Attention:
Rulemaking and Adjudications Staff.
Participants filing a document in this
manner are responsible for serving the
document on all other participants.
Filing is considered complete by firstclass mail as of the time of deposit in
the mail, or by courier, express mail, or
expedited delivery service upon
depositing the document with the
provider of the service. A presiding
officer, having granted an exemption
request from using E-Filing, may require
a participant or party to use E-Filing if
the presiding officer subsequently
determines that the reason for granting
the exemption from use of E-Filing no
longer exists.
Documents submitted in adjudicatory
proceedings will appear in the NRC’s
electronic hearing docket which is
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available to the public at https://
ehd1.nrc.gov/ehd/, unless excluded
pursuant to an order of the Commission,
or the presiding officer. Participants are
requested not to include personal
privacy information, such as social
security numbers, home addresses, or
home phone numbers in their filings,
unless an NRC regulation or other law
requires submission of such
information. However, a request to
intervene will require including
information on local residence in order
to demonstrate a proximity assertion of
interest in the proceeding. With respect
to copyrighted works, except for limited
excerpts that serve the purpose of the
adjudicatory filings and would
constitute a Fair Use application,
participants are requested not to include
copyrighted materials in their
submission.
Petitions for leave to intervene must
be filed no later than 60 days from the
date of publication of this notice.
Requests for hearing, petitions for leave
to intervene, and motions for leave to
file new or amended contentions that
are filed after the 60-day deadline will
not be entertained absent a
determination by the presiding officer
that the filing demonstrates good cause
by satisfying the following three factors
in 10 CFR 2.309(c)(1): (i) The
information upon which the filing is
based was not previously available; (ii)
the information upon which the filing is
based is materially different from
information previously available; and
(iii) the filing has been submitted in a
timely fashion based on the availability
of the subsequent information.
For further details with respect to this
license amendment application, see the
application for amendment, which is
available for public inspection at the
NRC’s PDR, located at One White Flint
North, Room O1–F21, 11555 Rockville
Pike (first floor), Rockville, Maryland
20852. Publicly available documents
created or received at the NRC are
accessible electronically through
ADAMS in the NRC Library at https://
www.nrc.gov/reading-rm/adams.html.
Persons who do not have access to
ADAMS or who encounter problems in
accessing the documents located in
ADAMS should contact the NRC’s PDR
Reference staff at 1–800–397–4209, 301–
415–4737, or by email to
pdr.resource@nrc.gov.
Detroit Edison, Docket No. 50–341,
Fermi 2, Monroe County, Michigan
Date of amendment request: January
11, 2013.
Description of amendment request:
The proposed amendment would revise
Fermi 2 Technical Specifications (TS) to
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incorporate the NRC-approved TSTF–
423, Revision 1. The proposed
amendment would modify TS to riskinform requirements regarding selected
Required Action end states by
incorporating the boiling water reactor
(BWR) owner’s group (BWROG)
approved Topical Report NEDC–32988–
A, Revision 2, ‘‘Technical Justification
to Support Risk-Informed Modification
to Selected Required Action End States
for BWR Plants.’’ Additionally, the
proposed amendment would modify the
TS Required Actions with a Note
prohibiting the use of limiting condition
for operation (LCO) 3.0.4.a when
entering the preferred end state (Mode
3) on startup.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Does the proposed change involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change allows a change to
certain required end states when the TS
Completion Times for remaining in power
operation will be exceeded. Most of the
requested technical specification (TS)
changes are to permit an end state of hot
shutdown (Mode 3) rather than an end state
of cold shutdown (Mode 4) contained in the
current TS. The request was limited to: (1)
Those end states where entry into the
shutdown mode is for a short interval, (2)
entry is initiated by inoperability of a single
train of equipment or a restriction on a plant
operational parameter, unless otherwise
stated in the applicable TS, and (3) the
primary purpose is to correct the initiating
condition and return to power operation as
soon as is practical. Risk insights from both
the qualitative and quantitative risk
assessments were used in specific TS
assessments. Such assessments are
documented in Section 6 of topical report
NEDC–32988–A, Revision 2, ‘‘Technical
Justification to Support Risk Informed
Modification to Selected Required Action
End States for BWR Plants.’’ They provide an
integrated discussion of deterministic and
probabilistic issues, focusing on specific TSs,
which are used to support the proposed TS
end state and associated restrictions. The
NRC staff finds that the risk insights support
the conclusions of the specific TS
assessments. Therefore, the probability of an
accident previously evaluated is not
significantly increased, if at all. The
consequences of an accident after adopting
TSTF–423 are no different than the
consequences of an accident prior to
adopting TSTF–423. Therefore, the
consequences of an accident previously
evaluated are not significantly affected by
this change. The addition of a requirement to
assess and manage the risk introduced by this
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change will further minimize possible
concerns.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed change does not involve a
physical alteration of the plant (no new or
different type of equipment will be installed).
If risk is assessed and managed, allowing a
change to certain required end states when
the TS Completion Times for remaining in
power operation are exceeded (i.e., entry into
hot shutdown rather than cold shutdown to
repair equipment) will not introduce new
failure modes or effects and will not, in the
absence of other unrelated failures, lead to an
accident whose consequences exceed the
consequences of accidents previously
evaluated. The addition of a requirement to
assess and manage the risk introduced by this
change and the commitment by the licensee
to adhere to the guidance in TSTF–IG–05–02,
‘‘Implementation Guidance for TSTF–423,
Revision 1, ‘Technical Specifications End
States, NEDC–32988–A,’’ will further
minimize possible concerns.
Thus, based on the above, this change does
not create the possibility of a new or different
kind of accident from an accident previously
evaluated.
3. Does the proposed change involve a
significant reduction in a margin of safety?
Response: No.
The proposed change allows, for some
systems, entry into hot shutdown rather than
cold shutdown to repair equipment, if risk is
assessed and managed. The BWROG’s risk
assessment approach is comprehensive and
follows NRC staff guidance as documented in
Regulatory Guides (RG) 1.174 and 1.177. In
addition, the analyses show that the criteria
of the three-tiered approach for allowing TS
changes are met. The risk impact of the
proposed TS changes was assessed following
the three-tiered approach recommended in
RG 1.177. A risk assessment was performed
to justify the proposed TS changes. The net
change to the margin of safety is
insignificant.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Bruce R.
Masters, DTE Energy, General Counsel—
Regulatory, 688 WCB, One Energy Plaza,
Detroit, MI 48226–1279.
NRC Branch Chief: Robert D. Carlson.
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Duke Energy Carolinas, LLC, Docket
Nos. 50–269, 50–270, and 50–287,
Oconee Nuclear Station, Units 1, 2, and
3 (ONS1, ONS2, and ONS3), Oconee
County, South Carolina
Date of amendment request: October
30, 2012.
Description of amendment request:
The proposed amendments would
revise the Technical Specifications
(TSs) to allow operation of a reverse
osmosis system during normal plant
operation to purify the water in the
borated water storage tanks and the
spent fuel pools.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Does the proposed change involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change requests NRC’s
approval of design features and controls that
will be used to ensure that periodic limited
operation of a Reverse Osmosis (RO) System
during Unit operation does not significantly
impact the Borated Water Storage Tank
(BWST) or Spent Fuel Pool (SFP) function or
other plant equipment. The proposed change
also requests NRC to approve proposed
Technical Specification (TS) requirements
that will impose operating restrictions and
isolation requirements on the RO System.
Duke Energy evaluated the effect of potential
failures, identified precautionary measures
that must be taken before and during RO
System operation, and identified specific
required operator actions to protect affected
structures, systems, and components (SSCs)
important to safety.
The new high energy piping and nonseismic piping being installed for the RO
System is non-QA1 and is postulated to fail
and cause an Auxiliary Building flood. Duke
Energy determined that adequate time is
available to isolate the flood source (BWST
or SFP) prior to affecting SSCs important to
safety.
The existing Auxiliary Building Flood
evaluation postulates a single break in the
non-seismic piping occurring in a seismic
event. The addition of the RO System will
not increase the probability of a seismic
event. The existing postulated source of the
pipe break in the Auxiliary Building is due
to the piping not being seismically designed.
The new RO System piping is considered a
potential source of a single pipe break for the
same reason. The new non-seismic RO
System piping is of similar quality as the
existing non-seismic piping and is no more
likely to fail than the existing piping. As
such, the addition of new non-seismic piping
does not significantly increase the probability
of occurrence of an Auxiliary Building flood
due to a single pipe break. An Auxiliary
Building flood due to a non-seismic RO
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System pipe break does not increase the
consequences of the flood since the new nonseismic pipe break is bounded by the
Auxiliary Building flood caused by existing
non-seismic pipe breaks.
Procedural controls will ensure that the
boron concentration does not go below the
TS limit as a result of water returned from
the RO System with lower boron
concentration. Thus, no adverse effects from
decreased boron concentration will occur.
The RO System takes suction from the top
of the SFP to protect SFP inventory. Plant
procedures will prohibit the use of the RO
System for the Units 1 & 2 SFP during the
time period directly after an outage that
requires the Units 1 & 2 SFP level to be
maintained higher than the TS Limiting
Condition for Operation (LCO) 3.7.11 level
requirement. The higher level is required to
support TS LCO 3.10.1 requirements for
Standby Shutdown Facility (SSF) Reactor
Coolant (RC) Makeup System operability
(due to the additional decay heat from the
recently offloaded spent fuel). Plant
procedures will also specify the siphon be
broken during this time period so the SFP
water above the RO suction point cannot be
siphoned off if the RO piping breaks. The
proposed change does not impact the fuel
assemblies, the movement of fuel, or the
movement of fuel shipping casks. The SFP
boron concentration, level, and temperature
limits will not be outside of required
parameters due to restrictions/requirements
on the system’s operation. In addition, the
proposed new TS will require the siphon be
broken during movement of irradiated fuel
assemblies in the SFP or movement of cask
over the SFP. Therefore, RO System
operation cannot occur during these
activities, effectively eliminating a Fuel
Handling Accidents (FHA) from occurring
while the RO System is in operation.
The BWST is used for mitigation of Steam
Generator Tube Rupture (SGTR), Main Steam
Line Break (MSLB), and Loss of Coolant
Accidents (LOCAs). The SGTR and MSLB are
bounded by the small break (SBLOCA)
analyses with respect to the performance
requirements for the High Pressure Injection
(HPI) System. In the normal mode of Unit
operation, the BWST is not an accident
initiator. The SFP is evaluated to maintain
acceptable criticality margin for all abnormal
and accident conditions including FHAs and
cask drop accidents. Both the BWST and SFP
are specified by TS requirements to have
minimum levels/volumes and boron
concentrations. The BWST also has TS
requirements for temperature. Prior to RO
System operation, procedures will require
the minimum required initial boron
concentration and initial level/volume to be
adjusted. Additionally, they will require the
RO System to be operated for a specified
maximum time period before readjusting
volume and boron concentration prior to
another RO session. This ensures that the TS
specified boron concentration and level/
volume limits for both the SFP and the
BWST are not exceeded during RO System
operation. Thus, the design functions of the
BWST and the SFP will continue to be met
during RO System operation.
Since the BWST and SFP will still have TS
boron concentration and level/volume
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requirements and the RO System will be
isolated prior to increasing radiation levels
preventing access to the isolation valve, the
mitigation of a LOCA or FHA does not result
in an increase in dose consequence. Since the
design basis LOCA analysis for Oconee
assumes 5 gpm back-leakage from the Reactor
Building sump to the BWST, the Emergency
Operating Procedure will require the RO
System to be isolated from the BWST prior
to switch over to the recirculation phase. The
proposed TS will require the RO system to
be isolated (by breaking the siphon) from the
SFPs during fuel handling activities and will
require the seismic boundary valve between
the BWST and RO System to be OPERABLE
in MODES 1, 2, 3, and 4.
The additional controls imposed by the
proposed Technical Specifications (TSs) will
provide additional assurance that isolation
valves and operating restrictions credited to
eliminate the need to analyze new release
pathways introduced by the RO system will
be in place.
Therefore, installation and operation of the
RO System during Unit operation and the
proposed TS imposing operating restrictions
do not significantly increase the probability
or consequences of any accident previously
evaluated.
2. Does the proposed change create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The RO System adds non-seismic piping in
the Auxiliary Building. However, the break of
a single non-seismic pipe in the Auxiliary
Building has already been postulated as an
event in the licensing basis. The RO System
also does not create the possibility of a
seismic event concurrent with a LOCA since
a seismic event is a natural phenomena
event. The RO System does not adversely
affect the Reactor Coolant System pressure
boundary. The suction to the RO System,
when using the system for BWST
purification, contains a normally closed
manual seismic boundary valve so the
seismic design criteria is met for separation
of seismic/non-seismic piping boundaries.
Duke Energy also evaluated potential
releases of radioactive liquid to the
environment due to RO System piping
failures. Design features, controls imposed by
the proposed TS, and procedural controls
will preclude release of radioactive materials
outside the Auxiliary Building by ensuring
the RO System will be isolated when
required.
The SFP suction line is designed such that
the SFP water level will not go below TS
required levels, thus the fuel assemblies will
have the TS required water level over them.
Procedural controls will restrict the use of
the RO System and require breaking vacuum
on the Units 1 & 2 SFP suction line when the
SSF conditions require the SFP level be
raised to support SSF RC Makeup System
operability. Thus, the SFP water level will
not be reduced below required water levels
for these conditions. RO System operating
restrictions will prevent reducing the SFP
boron concentration below TS limits.
Since the BWST and SFP will still have TS
boron concentration and level/volume
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requirements and the RO System will be
isolated prior to increasing radiation levels
preventing access to the isolation valve, the
mitigation of a LOCA or FHA does not result
in an increase in dose consequence. Since the
design basis LOCA analysis for Oconee
assumes 5 gpm back-leakage from the Reactor
Building sump to the BWST, the Emergency
Operating Procedure will require the RO
System to be isolated from the BWST prior
to switch over to the recirculation phase. The
proposed TS will require the RO system to
be isolated (by breaking the siphon) from the
SFPs prior to movement of irradiated fuel
assemblies in the SFP or movement of cask
over the SFP and will require the seismic
boundary valve between the BWST and RO
System to be OPERABLE in MODES 1, 2, 3,
and 4.
The additional controls imposed by the
proposed TSs will provide additional
assurance that isolation valves and operating
restrictions credited to eliminate the need to
analyze new release pathways introduced by
the RO system will be in place.
Therefore, operation of the RO System
during Unit operation will not create the
possibility of a new or different kind of
accident from any kind of accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The RO System adds non-seismic piping in
the Auxiliary Building. Duke Energy
evaluated the impact of RO System operation
on SSCs important to safety and determined
that the proposed TS controls and procedural
controls will ensure that TS limits for SFP
and BWST volume, temperature, and boron
concentration will continue to be met during
RO operation. For the BWST, these controls
will ensure the TS minimum BWST boron
concentration and level are available to
mitigate the consequences of a small break
LOCA or a large break LOCA. For the SFP,
these controls ensure the assumptions of the
fuel handling and cask drop accident
analyses are preserved. Additionally, the
failure of non-seismic RO System piping will
not significantly impact SSCs important to
safety. Oconee’s licensing basis does not
assume a design basis event occurs
simultaneously with a seismic event. The
proposed change does not significantly
impact the condition or performance of SSCs
relied upon for accident mitigation. This
change does not alter the existing TS
allowable values or analytical limits. The
existing operating margin between Unit
conditions and actual Unit setpoints is not
significantly reduced due to these changes.
The assumptions and results in any safety
analyses are not impacted. Therefore,
operation of the RO System during Unit
operation does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
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Attorney for licensee: Lara S. Nichols,
Associate General Counsel, Duke Energy
Corporation, 526 South Church Street–
EC07H, Charlotte, NC 28202–1802.
NRC Branch Chief: Robert J.
Pascarelli.
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Duke Energy Carolinas, LLC, Docket
Nos. 50–269, 50–270, and 50–287,
Oconee Nuclear Station, Units 1, 2, and
3, Oconee County, South Carolina
Date of amendment request: February
22, 2013.
Description of amendment request:
The proposed amendments would
revise the Technical Specification
curves for pressure and temperature
limits on the reactor coolant system, and
limits on heatup and cooldown rates.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed amendment replaces the
current Oconee Nuclear Station (ONS) Units
1, 2, and 3 pressure/temperature (P–T) limit
curves applicable to 33 effective full power
years (EFPY) in Technical Specification (TS)
3.4.3 with new P–T limit curves applicable
to 54 EFPY. The proposed changes also
revise the Reactor Coolant System heatup
and cooldown rates and allowable reactor
coolant pump combinations of TS Tables
3.4.3–1 and 3.4.3–2. The pressuretemperature (P–T) limit curves in the TSs
were conservatively generated in accordance
with fracture toughness requirements of
ASME Code Section XI, Appendix G, and the
minimum pressure and temperature
requirements as listed in Table 1 of 10 CFR
Part 50, Appendix G. The proposed changes
do not impact the capability of the reactor
coolant pressure boundary (i.e., no change in
operating pressure, materials, seismic
loading, etc.).
Therefore, the proposed changes do not
increase the potential for the occurrence of a
loss of coolant accident (LOCA). The changes
do not modify the reactor coolant system
pressure boundary, nor make any physical
changes to the facility design, material, or
construction standards. The probability of
any design basis accident (DBA) is not
affected by this change, nor are the
consequences of any DBA affected by this
change. The proposed P–T limits, heatup and
cooldown rates and allowable operating
reactor coolant pump combinations are not
considered to be an initiator or contributor to
any accident analysis addressed in the ONS
Updated Final Safety Analyses Report
(UFSAR).
The proposed changes will not impact
assumptions and conditions previously used
in the radiological consequence evaluations
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nor affect the mitigation of these
consequences due to an accident described in
the UFSAR. Also, the proposed changes will
not impact a plant system such that
previously analyzed SSCs might be more
likely to fail. The initiating conditions and
assumptions for accidents described in the
UFSAR remain as analyzed.
Therefore, the probability or consequences
of an accident previously evaluated is not
significantly increased.
2. Does the proposed amendment create
the possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The requirements for P–T limit curves have
been in place since the beginning of plant
operation. The revised curves are based on a
later edition to Section XI of the ASME Code
that incorporates current industry standards
for P–T curves. The revised curves are based
on reactor vessel irradiation damage
predictions using Regulatory Guide 1.99
methodology. No new failure modes are
identified nor are any SSCs required to be
operated outside the design bases.
Therefore, the possibility of a new or
different kind of accident from any kind of
accident previously evaluated is not created.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed P–T curves continue to
maintain the safety margins of 10 CFR Part
50, Appendix G, by defining the limits of
operation which prevent non-ductile failure
of the reactor pressure vessel. Analyses have
demonstrated that the fracture toughness
requirements are satisfied and that
conservative operating restrictions are
maintained for the purpose of low
temperature overpressure protection. The P–
T limit curves provide assurance that the
RCS pressure boundary will behave in a
ductile manner and that the probability of a
rapidly propagating fracture is minimized.
Therefore, this request does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Lara S. Nichols,
Deputy General Counsel, Duke Energy
Corporation, 526 South Church Street–
EC07H, Charlotte, NC 28202–1802.
NRC Branch Chief: Robert J.
Pascarelli.
Exelon Generation Company, LLC,
Docket Nos. STN 50–456 and STN 50–
457, Braidwood Station, Units 1 and 2,
Will County, Illinois
Docket Nos. STN 50–454 and STN 50–
455, Byron Station, Units 1 and 2, Ogle
County, Illinois
Date of amendment request:
December 21, 2012.
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Description of amendment request:
The proposed amendment would Revise
Technical Specifications (TS) 3.3.6,
‘‘Containment Ventilation Isolation
Instrumentation.’’ Specifically, this
amendment request proposes to revise
Footnote (b) of TS Table 3.3.6–1,
‘‘Containment Ventilation Isolation
Instrumentation,’’ which specifies the
‘‘Containment Radiation—High’’ trip
setpoint for two containment area
radiation monitors (i.e., 1(2) RE–AR011
and 1(2) RE–AR012). The proposed
changes would revise the ‘‘Containment
Radiation—High’’ trip setpoint from the
current, overly conservative value (i.e.,
a submersion dose rate of less than or
equal to 10 mRhr in the containment
building), to less than or equal to 2
times the containment building
background radiation reading at rated
thermal power, which is consistent with
NUREG–1431, ‘‘Standard Technical
Specifications, Westinghouse Plants.’’
Upon reaching the ‘‘Containment
Radiation—High’’ setpoint, these area
radiation monitors provide an isolation
signal to the containment normal purge,
mini purge and post-LOCA (Loss of
Coolant Accident) systems’ containment
isolation valves.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Does the proposed change involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The containment ventilation isolation
radiation monitors serve two primary
functions, they:
a. act as backup to the SI [safety injection]
signal to ensure closing of the purge valves;
and
b. are the primary means for automatically
isolating containment in the event of a fuel
handling accident in containment.
Upon sensing a high radiation condition in
containment, these area radiation monitors
provide an isolation signal to the
containment normal purge, mini purge and
post- LOCA systems containment isolation
valves (i.e., a containment ventilation
isolation signal).
The accidents that could potentially be
impacted by the proposed change were
evaluated; specifically the Loss of Coolant
Accident (LOCA), Control Rod Ejection
Accident (CREA) and Fuel Handling
Accident (FHA) in Containment. The
proposed change has no impact on
probability of these accidents occurring as
the subject containment radiation area
monitors detect a high radiation condition
resulting from these accidents. The radiation
monitors do not initiate any accidents or
transients. Changing the ‘‘Containment
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Radiation—High’’ trip setpoint from ‘‘≤10
mR/hr in the containment building,’’ to ‘‘≤2
times the containment building background
radiation reading at rated thermal power’’
only affects the point (i.e., the radiation level
in containment) at which a containment
ventilation isolation signal would be
generated. The requested change does not
involve any physical plant modifications or
operational changes that could adversely
affect system reliability or performance of the
radiation monitors, or that could affect the
probability of operator error.
The requested change does not affect any
postulated accident precursors and therefore,
will not affect the probability of an accident
previously evaluated.
The proposed change was evaluated to
determine the impact on the dose
consequences of the LOCA, CREA, or FHA in
containment. The evaluation assumed a
containment purge was in progress at the
onset of the subject accidents and showed
that the proposed change in the containment
radiation monitors’ setpoint had no effect on
the purge valve isolation time. With regard to
the LOCA and CREA, the safety analysis
assumes a prompt purge valve isolation time
(i.e., approximately 60 seconds) that
significantly bounds the radiation monitor
sensing and response time, and actual valve
design closure time (i.e., a total of
approximately 7 seconds). The radiation
monitor setpoint is not considered in the
safety analysis and any dose contribution
associated with the containment purge, due
to the proposed change in setpoint, was
shown to be unaffected; therefore, the
proposed change has no impact on the
already insignificant dose contribution
attributed to a containment purge during an
accident of less than one mrem.
The dose consequences associated with the
FHA in containment are also not impacted by
the proposed change in containment
radiation monitor setpoint. The existing dose
consequences resulting from a FHA with
moving non-RECENTLY IRRADIATED FUEL
(i.e., fuel moved more than 48 hours after
reactor shutdown) conservatively assume the
containment purge valves remain open
throughout the event; therefore, a change in
the isolation setpoint does not impact the
results of this analysis. With regard to
movement of RECENTLY IRRADIATED
FUEL (i.e., fuel moved less then 48 hours
after reactor shutdown), EGC’s [Exelon
Generation Company] proposal deletes TS
LCO [limiting condition for operation]
3.9.4.c.2 which allowed the containment
purge valves to be open provided the
containment radiation isolation system is
OPERABLE. Deletion of TS LCO 3.9.4.c.2
ensures that the containment purge valves
are in the closed position when moving
RECENTLY IRRADIATED FUEL, thus
removing dependence on the containment
radiation isolation system and associated
radiation monitor setpoint from the FHA
dose consequences.
The four other additional TS changes
associated with the deletion of LCO 3.9.4,
Item c.2, proposed for consistency (i.e.,
deleting a NOTE regarding MODE
applicability, deleting a CONDITION related
only to LCO 3.9.4.c.2, deleting a footnote
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regarding MODE applicability; and deleting
two surveillances related to LCO 3.9.4.c.2),
also have no affect on either the probability
or consequences of an accident previously
evaluated.
Based on the above discussion, the
proposed change does not involve a
significant increase in the probability or
consequences of an accident previously
evaluated.
2. Does the proposed change create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed changes do not result in a
change to the design of the Containment
Ventilation Isolation System or the manner in
which the system operates or provides plant
protection. The containment radiation
monitors will sense radiation levels in the
same way and will respond in the same
manner when the setpoint is exceeded. The
change in the ‘‘Containment Radiation—
High’’ setpoint does not create a new failure
mode for the associated containment
radiation monitors or for any other plant
equipment. The deletion of LCO 3.9.4, Item
c.2, in support of the setpoint change during
refueling operations, is more conservative
than the current allowances and actually
eliminates a potential failure mode for the
assumed open containment ventilation
isolation valves as the proposed deletion of
LCO 3.9.4, Item c.2 would require the valves
to be closed prior to moving RECENTLY
IRRADIATED FUEL.
The changes do not result in the creation
of any new accident precursors, the creation
of any changes to the existing accident
scenarios, nor do they create any new or
different accident scenarios. Subsequently,
the accidents defined in the UFSAR [updated
final safety analysis report] continue to
represent the credible spectrum of events to
be analyzed which determine safe plant
operation.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any previously
evaluated.
3. Does the proposed change involve a
significant reduction in a margin of safety?
Response: No.
The analysis methodologies used in the
subject safety analyses are not modified as a
result of the proposed TS changes to the
‘‘Containment Radiation—High’’ trip setpoint
or the deletion of LCO 3.9.4, Item c.2, or any
of the other four associated TS changes.
Although the ‘‘Containment Radiation—
High’’ trip setpoint is being increased, the
increase in response time to a high radiation
condition in containment, when compared to
the current setpoint, is negligible due to the
projected prompt rise in containment
radiation level upon initiation of a LOCA.
The dose consequences and resultant margin
of safety to the regulatory acceptance limits,
due to revising the ‘‘Containment
Radiation—High’’ setpoint to ≤ 2 times the
containment building background radiation
reading at rated thermal power, was shown
to be unaffected for normal at-power
containment releases; have a negligible
impact on the associated LOCA and CREA
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accident dose consequences; and have no
impact on the FHA when moving RECENTLY
IRRADIATED FUEL. Therefore, the proposed
changes do not impact any analysis margins.
The proposed changes do not alter the
manner in which the safety limits, limiting
safety system setpoints, or limiting
conditions for operation are determined. The
current safety analyses remain bounding
since their conclusions are not affected by
the proposed changes. The safety systems
credited in the safety analyses will continue
to be available to perform their mitigation
functions. All protection signals credited as
the primary or secondary accident mitigating
functions, and all operator actions credited in
the accident analyses remain the same. The
proposed changes will not result in plant
operation in a configuration outside the
design basis.
Based on the above information, the
proposed change does not result in a
significant reduction in the margin of safety.
Based on the above evaluation, EGC
concludes that the proposed amendments do
not involve a significant hazards
consideration under the standards set forth in
10 CFR 50.92, paragraph (c), and,
accordingly, a finding of no significant
hazards consideration is justified.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
requested amendments involve no
significant hazards consideration.
Attorney for licensee: Mr. Bradley J.
Fewell, Associate General Counsel,
Exelon Nuclear, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Acting Branch Chief: Jeremy S.
Bowen.
Florida Power and Light Company,
Docket Nos. 50–250 and 50–251, Turkey
Point Nuclear Generating Units 3 and 4,
Miami-Dade County, Florida
Date of amendment request: January
29, 2013.
Description of amendment request:
The license amendment request
proposes to remove completed and
satisfied license conditions and to
correct inadvertent errors and incorrect
references.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed amendments do not change
or modify the fuel, fuel handling processes,
fuel storage racks, number of fuel assemblies
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that may be stored in the spent fuel pool
(SFP), decay heat generation rate, or the
spent fuel pool cooling and cleanup system.
The proposed amendments only limit
crediting of burnable absorbers in the spent
fuel pool to Integrated Fuel Burnable
Absorber (IFBA) rods that were specifically
addressed in the currently approved
criticality analysis ([Westinghouse
Commercial Atomic Power report] WCAP–1
7094–P, Revision 3). The removal of the
phrase ‘‘or an equivalent amount of another
burnable absorber’’ eliminates the possibility
of crediting a burnable absorber other than
IFBA for storage of spent fuel assemblies in
the spent fuel pool without prior NRC’s
approval. The deletion of the license
condition associated with the Boraflex
Remedy is editorial as it is no longer
applicable. The proposed amendments do
not affect the ability of the BAST [boric acid
storage tank] to perform its function or the
ability of the CREVS [control room
emergency ventilation system] to perform its
function. These latter proposed TS [technical
specification] changes correct inadvertent
errors and are consistent with the stated
intent of original license submittals or delete
license conditions that are no longer
applicable or that have been fully satisfied.
The proposed amendments do not cause
any physical change to the existing spent fuel
storage configuration, fuel makeup, RCS
[reactor coolant system] pressure boundary,
reactor containment, or plant systems. The
proposed amendments do not affect any
precursors to any accident previously
evaluated or do not affect any known
mitigation equipment or strategies.
Therefore, the proposed changes do not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create
the possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed amendments do not change
or modify the fuel, fuel handling processes,
fuel racks, number of fuel assemblies that
may be stored in the pool, decay heat
generation rate, or the spent fuel pool cooling
and cleanup system. The proposed
amendments do not result in any changes to
spent fuel or to fuel storage configurations.
The removal of the phrase ‘‘or an equivalent
amount of another burnable absorber’’
eliminates the possibility of crediting a
burnable absorber other than IFBA for storage
of spent fuel assemblies in the spent fuel
pool without prior NRC approval. The
proposed amendments do not affect the
ability of the BAST to perform its function
or the ability of the CREVS to perform its
function. These latter proposed TS changes
correct inadvertent errors and are consistent
with the stated intent of the original license
submittals, delete license conditions that are
no longer applicable or have been fully
satisfied.
Therefore, the proposed changes do not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
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3. Does the proposed amendment involve
a significant reduction in the margin of
safety?
Response: No.
The proposed amendments do not change
or modify the fuel, fuel handling processes,
fuel racks, number of fuel assemblies that
may be stored in the pool, decay heat
generation rate, or the spent fuel pool cooling
and cleanup system. Therefore, the proposed
amendments have no impact to the existing
margin of safety for subcriticality required by
10 CFR 50.68(b)(4). The other proposed OL
[operating license] & TS changes correct
inadvertent errors and are consistent with the
stated intent of the original license submittals
or delete license conditions that are no longer
applicable or have been fully satisfied.
Therefore, the proposed amendments do
not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 50.92(c) are satisfied.
Therefore, the NRC staff proposes to
determine that the amendment request
involves no significant hazards
consideration.
Attorney for licensee: James Petro,
Managing Attorney—Nuclear, Florida
Power & Light, P.O. Box 14000, Juno
Beach, Florida 33408–0420.
NRC Branch Chief: Jessie F.
Quichocho.
Nebraska Public Power District, Docket
No. 50–298, Cooper Nuclear Station,
Nemaha County, Nebraska
Date of amendment request: June 25,
2012.
Description of amendment request:
The amendment would revise the
description of the Fuel Handling
Accident (FHA) in Section XIV–6.4 of
the Cooper Nuclear Station (CNS)
Updated Safety Analysis Report
(USAR). The revised USAR FHA
description is based on changes to the
Design Basis Accident FHA dose
calculation, to reflect a 24-month fuel
cycle source term using a Global
Nuclear Fuels (GNF) 10 × 10 fuel array,
reduce the bounding Radial Peaking
Factor, and revise the total effective
dose equivalent (TEDE) contribution to
consider the shine contribution from
external sources.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Do the proposed changes involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
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The analyses changes described by this
proposed change to the USAR are not
initiators to events, and, therefore, do not
involve the probability of an accident. The
changes to the FHA calculation for
radiological dose following a FHA
incorporate the following:
—accounts for the increase to the source term
owing to the use of Global Nuclear Fuels
(GNF) 10 × 10 fuel exposed over a 24month fuel cycle,
—reduces the Radial Peaking Factor from 2.0
to 1.95, and
—uses a calculated Control Room shine
contribution that is added to the FHA dose
consequences.
The NRC computer code RADTRAD
Version 3.03 is used for the offsite and
Control Room dose calculation. The
RADTRAD code was approved for use with
the CNS FHA alternative source term (AST)
dose calculation in License Amendment 222.
Because the analysis affected by the
changes are not considered to be an initiator
to any previously analyzed accident, these
changes cannot increase the probability of
any previously evaluated accident. Therefore,
these changes do not increase the probability
of occurrence of an accident evaluated
previously in the USAR.
The changes in FHA dose consequences to
the Control Room occupant resulting from
the 24-month cycle/GNF 10 × 10 source term
(without crediting the offset by a reduced
Radial Peaking Factor), results in more than
a minimal increase in the consequences of an
accident previously evaluated in the USAR,
as stated in 10 CFR 50.59(c)(2)(iii). However,
the resultant dose remains well within the
regulatory limits of 10 CFR 50.67. When the
reduced Radial Peaking Factor is applied, the
dose consequences are minor. Therefore, this
change does not significantly increase the
consequences of an accident previously
evaluated in the USAR.
In summary, the proposed change does not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
This change does not involve initiators to
any events in the USAR, nor does the activity
create the possibility for any new accidents.
Rather, this change is a result of the
evaluation of the most limiting FHA, which
can occur at CNS. The changes to the FHA
calculation for radiological dose following a
FHA incorporate the following:
—accounts for the increase to the source term
owing to the use of GNF 10 × 10 fuel
exposed over a 24-month fuel cycle,
—reduces the Radial Peaking Factor from 2.0
to 1.95, and
—uses a calculated Control Room shine
contribution that is added to the FHA dose
consequences.
The RADTRAD code accommodates the
use of GNF 10 × 10 fuel exposed over a 24month fuel cycle in calculating the FHA dose
consequences. The reduction in Radial
Peaking Factor remains bounding over the
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CNS core design. The calculated Control
Room shine contribution replaces the
previously approved qualitative assessment.
The proposed change does not introduce any
new modes of plant operation and does not
involve physical modifications to the plant.
As a result, no new failure modes are being
introduced.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any previously
evaluated.
3. Do the proposed changes involve a
significant reduction in a margin of safety?
Response: No.
The dose consequences are calculated in
accordance with the regulatory guidance
found in RG 1.183. The RADTRAD code was
used, as approved for application at CNS
with License Amendment 222. With the
reduced Radial Peaking Factor applied to the
GNF 10 × 10 fuel that has been exposed over
a 24-month fuel cycle, the dose consequences
are minor. The calculated shine contribution
being added to the total Control Room
occupant FHA dose results are less than the
previous qualitative assessment results that
are being replaced. Accordingly, the safety
margins to the regulatory dose limits are
preserved.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
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The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Mr. John C.
McClure, Nebraska Public Power
District, Post Office Box 499, Columbus,
NE 68602–0499.
NRC Branch Chief: Michael T.
Markley.
NextEra Energy Duane Arnold, LLC,
Docket No. 50–331, Duane Arnold
Energy Center, Linn County, Iowa
Date of amendment request:
November 13, 2012.
Description of amendment request:
The proposed amendment would revise
Renewed Facility Operating License
(RFOL) Condition C.12 to clarify that
the programs and activities, identified
in Appendix A of NUREG–1955 and the
Updated Final Safety Analysis Report
(UFSAR) supplement are to be
completed before the period of extended
operation.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Does the proposed amendment involve
a significant increase in the probability or
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consequences of an accident previously
evaluated?
Response: No.
The amendment does not significantly
increase the probability of an accident since
it does not involve a change to any plant
equipment that initiates a plant accident. The
change clarifies RFOLC [RFOL Condition]
C.12. The license conditions deal with
administrative controls over information
contained in the Updated Final Safety
Analysis Repo[r]t (UFSAR) supplement. The
proposed changes are administrative and the
license conditions are not an initiator or
mitigator of any previously evaluated
accidents.
Therefore, the proposed amendment does
not involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create
the possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed amendment does not create
the possibility of a new or different kind of
accident from any accident previously
evaluated since it does not involve any
physical alteration of plant equipment and
does not change the method by which any
safety-related system performs its function.
The license conditions deal with
administrative controls over information
contained in the UFSAR supplement. No
new or different types of equipment will be
installed and the basic operation of installed
equipment is unchanged.
Therefore, the proposed amendment does
not create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed amendment does not affect
design codes or design margins. The change
that clarifies RFOLC C.12 is administrative in
nature and does not have the ability to affect
analyzed safety margins.
Therefore, operation of DAEC in
accordance with the proposed amendment
will not involve a significant reduction in the
margin to safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Mr. James Petro,
P. O. Box 14000, Juno Beach, FL 33408–
0420.
NRC Branch Chief: Robert D. Carlson.
NextEra Energy Duane Arnold, LLC,
Docket No. 50–331, Duane Arnold
Energy Center, Linn County, Iowa
Date of amendment request:
December 21, 2012.
Description of amendment request:
The proposed amendment would
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modify the current DAEC Technical
Specifications (TS) requirement to
operate the Standby Gas Treatment
System for 10 hours on a frequency
specified in the Surveillance Frequency
Control Program in accordance with
TSTF–522, Revision 0, ‘‘Revise
Ventilation System Surveillance
Requirements to Operate for 10 hours
per Month.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Does the proposed change involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change replaces an existing
Surveillance Requirement to operate the SGT
System equipped with electric heaters for a
continuous 10 hour period with a
requirement to operate the SGT System for 15
continuous minutes without the heaters
operating. In addition, the electrical heater
output test in the VFTP (Specification
5.5.7.e) is proposed to be deleted and a
corresponding change in the charcoal filter
testing (Specification 5.5.7.c) be made to
require the testing be conducted at a
humidity of at least 95% RH, which is more
stringent than the current testing requirement
of 70% RH.
These systems are not accident initiators
and therefore, these changes do not involve
a significant increase in the probability of an
accident. The proposed system and filter
testing changes are consistent with current
regulatory guidance for these systems and
will continue to assure that these systems
perform their design function which may
include mitigating accidents. Thus the
change does not involve a significant
increase in the consequences of an accident.
Therefore, it is concluded that this change
does not involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed change replaces an existing
Surveillance Requirement to operate the SGT
System equipped with electric heaters for a
continuous 10 hour period with a
requirement to operate the systems for 15
continuous minutes without the heaters
operating. In addition, the electrical heater
output test in the VFTP (Specification
5.5.7.e) is proposed to be deleted and a
corresponding change in the charcoal filter
testing (Specification 5.5.7.c) be made to
require the testing be conducted at a
humidity of at least 95% RH, which is more
stringent than the current testing requirement
of 70% RH.
The change proposed for this ventilation
system does not change any system
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operations or maintenance activities. Testing
requirements will be revised and will
continue to demonstrate that the Limiting
Conditions for Operation are met and the
system components are capable of
performing their intended safety functions.
The change does not create new failure
modes or mechanisms and no new accident
precursors are generated.
Therefore, it is concluded that this change
does not create the possibility of a new or
different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a
significant reduction in a margin of safety?
Response: No.
The proposed change replaces an existing
Surveillance Requirement to operate the SGT
System equipped with electric heaters for a
continuous 10 hour period with a
requirement to operate the systems for 15
continuous minutes without heaters
operating. In addition, the electrical heater
output test in the VFTP is proposed to be
deleted and a corresponding change in the
charcoal filter testing be made to require the
testing be conducted at a humidity of at least
95% RH, which is more stringent than the
current testing requirement of 70% RH.
The proposed increase to 95% RH in the
required testing of the charcoal filters
compensates for the function of the heaters,
which was to reduce the humidity of the
incoming air to below the currently-specified
value of 70% RH for the charcoal. The
proposed change is consistent with
regulatory guidance and continues to ensure
that the performance of the charcoal filters is
acceptable.
Therefore, it is concluded that this change
does not involve a significant reduction in a
margin of safety.
mstockstill on DSK4VPTVN1PROD with NOTICES
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Mr. James Petro,
P.O. Box 14000, Juno Beach, FL 33408–
0420.
NRC Branch Chief: Robert D. Carlson.
R.E. Ginna Nuclear Power Plant, LLC,
Docket No. 50–244, R.E. Ginna Nuclear
Power Plant, Wayne County, New York
Date of amendment request: April 20,
2012.
Description of amendment request:
The proposed amendment would revise
the TS 3.1.7 to approve the use of an
alternative method, other than the
current method of the use of movable
incore detectors system, to monitor the
position of control rod or shutdown rod,
in the event of a malfunction of the
microprocessor rod position indication
(MRPI) system. The use of this
alternative method would reduce the
required frequency of flux mapping
using the movable incore detector
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system to determine the position of the
control or shutdown rod position that is
not being indicated. This will reduce
the wear on the movable incore detector
system that is also used to complete
other required TS surveillances.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Does the proposed change involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change provides an
alternative method for verifying rod position
of one rod. The proposed change meets the
intent of the current specification in that it
ensures verification of position of the rod
once every 8 hours. The proposed change
provides only an alternative method of
monitoring rod position and does not change
the assumptions or results of any previously
evaluated accident.
Therefore, operation of the facility in
accordance with the proposed amendment
would not involve a significant increase in
the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed change provides only an
alternative method of determining the
position of one rod. No new accident
initiators are introduced by the proposed
alternative manner of performing rod
position verification. The proposed change
does not affect the reactor protection system.
Hence, no new failure modes are created that
would cause a new or different kind of
accidents from any accident previously
evaluated.
Therefore, operation of the facility in
accordance with the proposed amendments
would not create the possibility of a new or
different kind of accident from any
previously evaluated.
3. Does the proposed change involve a
significant reduction in a margin of safety?
Response: No.
The basis of TS 3.1.7 states that the
operability of the rod position indicators is
required to determine control rod positions
and thereby ensure compliance with the
control rod alignment and insertion limits.
The proposed change does not alter the
requirement to determine rod position but
provides an alternative method for
determining the position of the affected rod.
As a result, the initial conditions of the
accident analysis are preserved and the
consequences of previously analyzed
accidents are unaffected.
Therefore, operation of the facility in
accordance with the proposed amendment
would not involve a significant reduction in
a margin of safety.
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The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Carey Fleming,
Sr. Counsel—Nuclear Generation,
Constellation Group, LLC, 750 East Pratt
Street, 17 Floor, Baltimore, MD 21202.
NRC Acting Branch Chief: Sean
Meighan.
South Carolina Electric and Gas Docket
Nos.: 52–027 and 52–028, Virgil C.
Summer Nuclear Station (VCSNS) Units
2 and 3, Fairfield County, South
Carolina
Date of amendment request: March
26, 2013.
Description of amendment request:
The proposed change would amend
Combined License Nos.: NPF–93 and
NPF–94 for Virgil C. Summer Nuclear
Station (VCSNS) Units 2 and 3 by
departing from the plant-specific design
control document Tier 2* material
contained within the Updated Safety
Analysis Report (UFSAR) by revising
the structural criteria code for anchoring
of reinforcement bar within the nuclear
island walls.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The design functions of the nuclear island
structures are to provide support, protection,
and separation for the seismic Category I
mechanical and electrical equipment located
in the nuclear island. The nuclear island
structures are structurally designed to meet
seismic Category I requirements as defined in
Regulatory Guide 1.29.
The change of the requirements for
anchoring headed reinforcement does not
have an adverse impact on the response of
the nuclear island structures to safe
shutdown earthquake ground motions or
loads due to anticipated transients or
postulated accident conditions. The change
of the requirements for anchoring headed
reinforcement does not impact the support,
design, or operation of mechanical and fluid
systems. There is no change to plant systems
or the response of systems to postulated
accident conditions. There is no change to
the predicted radioactive releases due to
postulated accident conditions. The plant
response to previously evaluated accidents or
external events is not adversely affected, nor
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does the change described create any new
accident precursors.
Therefore, the proposed amendment does
not involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create
the possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed change is to provide the
requirements for anchoring nuclear island
headed reinforcement. The proposed change
does not change the design of the nuclear
island structures except to a limited extent to
redistribute the shear reinforcement in the
walls of the nuclear island. The proposed
change does not impact the support, design,
or operation of mechanical or fluid systems.
The proposed change does not result in a
new failure mechanism for the nuclear island
structures or new accident precursors. As a
result, the design functions of the nuclear
island structures and the seismic Category I
mechanical and electrical equipment located
in the nuclear island are not adversely
affected by the proposed change.
Therefore, the proposed amendment does
not create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
No safety analysis or design basis
acceptance limit/criterion is challenged or
exceeded by the proposed change, thus, no
margin of safety is reduced. The limited
application of alternative criteria for headed
reinforcement does not reduce the margin of
safety.
Therefore, the proposed amendment does
not involve a significant reduction in a
margin of safety.
mstockstill on DSK4VPTVN1PROD with NOTICES
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Ms. Kathryn M.
Sutton, Morgan, Lewis & Bockius LLC,
1111 Pennsylvania Avenue NW.,
Washington, DC 20004–2514.
NRC Branch Chief: Lawrence
Burkhart, Acting.
Southern Nuclear Operating Company
Docket Nos.: 52–025 and 52–026, Vogtle
Electric Generating Plant (VEGP) Units
3 and 4, Burke County, Georgia
Date of amendment request: March
20, 2013.
Description of amendment request:
The proposed change would amend
Combined Licenses Nos.: NPF–91 and
NPF–92 for Vogtle Electric Generating
Plant (VEGP) Units 3 and 4 by departing
from the plant-specific design control
document Tier 2* material contained
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22573
within the Updated Safety Analysis
Report (UFSAR) by revising the
structural criteria code for anchoring of
reinforcement bar within the nuclear
island walls.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
No safety analysis or design basis
acceptance limit/criterion is challenged or
exceeded by the proposed change, thus, no
margin of safety is reduced. The limited
application of alternative criteria for headed
reinforcement does not reduce the margin of
safety.
Therefore, the proposed amendment does
not involve a significant reduction in a
margin of safety.
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The design functions of the nuclear island
structures are to provide support, protection,
and separation for the seismic Category I
mechanical and electrical equipment located
in the nuclear island. The nuclear island
structures are structurally designed to meet
seismic Category I requirements as defined in
Regulatory Guide 1.29.
The change of the requirements for
anchoring headed reinforcement does not
have an adverse impact on the response of
the nuclear island structures to safe
shutdown earthquake ground motions or
loads due to anticipated transients or
postulated accident conditions. The change
of the requirements for anchoring headed
reinforcement does not impact the support,
design, or operation of mechanical and fluid
systems. There is no change to plant systems
or the response of systems to postulated
accident conditions. There is no change to
the predicted radioactive releases due to
postulated accident conditions. The plant
response to previously evaluated accidents or
external events is not adversely affected, nor
does the change described create any new
accident precursors.
Therefore, the proposed amendment does
not involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create
the possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed change is to provide the
requirements for anchoring nuclear island
headed reinforcement. The proposed change
does not change the design of the nuclear
island structures except to a limited extent to
redistribute the shear reinforcement in the
walls of the nuclear island. The proposed
change does not impact the support, design,
or operation of mechanical or fluid systems.
The proposed change does not result in a
new failure mechanism for the nuclear island
structures or new accident precursors. As a
result, the design functions of the nuclear
island structures and the seismic Category I
mechanical and electrical equipment located
in the nuclear island are not adversely
affected by the proposed change.
Therefore, the proposed amendment does
not create the possibility of a new or different
kind of accident from any accident
previously evaluated.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Mr. M. Stanford
Blanton, Blach & Bingham LLP, 1710
Sixth Avenue North, Birmingham, AL
35203–2015.
NRC Acting Branch Chief: Lawrence
Burkhart.
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Notice of Issuance of Amendments to
Facility Operating Licenses and
Combined Licenses
During the period since publication of
the last biweekly notice, the
Commission has issued the following
amendments. The Commission has
determined for each of these
amendments that the application
complies with the standards and
requirements of the Atomic Energy Act
of 1954, as amended (the Act), and the
Commission’s rules and regulations.
The Commission has made appropriate
findings as required by the Act and the
Commission’s rules and regulations in
10 CFR Chapter I, which are set forth in
the license amendment.
A notice of consideration of issuance
of amendment to facility operating
license or combined license, as
applicable, proposed no significant
hazards consideration determination,
and opportunity for a hearing in
connection with these actions, was
published in the Federal Register as
indicated.
Unless otherwise indicated, the
Commission has determined that these
amendments satisfy the criteria for
categorical exclusion in accordance
with 10 CFR 51.22. Therefore, pursuant
to 10 CFR 51.22(b), no environmental
impact statement or environmental
assessment need be prepared for these
amendments. If the Commission has
prepared an environmental assessment
under the special circumstances
provision in 10 CFR 51.22(b) and has
made a determination based on that
assessment, it is so indicated.
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For further details with respect to the
action see (1) the applications for
amendment, (2) the amendment, and (3)
the Commission’s related letter, Safety
Evaluation and/or Environmental
Assessment as indicated. All of these
items are available for public inspection
at the NRC’s Public Document Room
(PDR), located at One White Flint North,
Room O1–F21, 11555 Rockville Pike
(first floor), Rockville, Maryland 20852.
Publicly available documents created or
received at the NRC are accessible
electronically through the Agencywide
Documents Access and Management
System (ADAMS) in the NRC Library at
https://www.nrc.gov/reading-rm/
adams.html. If you do not have access
to ADAMS or if there are problems in
accessing the documents located in
ADAMS, contact the PDR’s Reference
staff at 1–800–397–4209, 301–415–4737
or by email to pdr.resource@nrc.gov.
mstockstill on DSK4VPTVN1PROD with NOTICES
Dominion Nuclear Connecticut, Inc.,
Docket No. 50–336, Millstone Power
Station, Unit 2, New London County,
Connecticut
Date of amendment request:
December 17, 2012, as supplemented by
January 31, 2013.
Description of amendment request:
The amendment revised the Millstone
Power Station, Unit 2 (MPS2) Technical
Specification (TS) Surveillance
Requirement 4.4.3.2 to remove the
requirement to perform the quarterly
surveillance for a pressurizer poweroperated relief valve (PORV) block valve
that is being maintained closed in
accordance with TS 3.4.3 Action a. The
proposed change is consistent with the
requirements of the Standard Technical
Specification—Combustion Engineering
Plants (NUREG–1432, Revision 4).
Date of issuance: March 26, 2013.
Effective date: As of the date of
issuance, and shall be implemented
within 60 days.
Amendment No.: 314.
Renewed Facility Operating License
No. DPR–65: Amendment revised the
License and Technical Specifications.
Date of initial notice in Federal
Register: January 22, 2012 (78 FR 4472).
The supplemental letter dated January
31, 2013, provided additional
information that clarified the
application, did not expand the scope of
the application as originally noticed,
and did not change the staff’s original
proposed no significant hazards
consideration determination.
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated March 26, 2013.
No significant hazards consideration
comments received: No.
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Entergy Gulf States Louisiana, LLC, and
Entergy Operations, Inc., Docket No.
50–458, River Bend Station, Unit 1,
West Feliciana Parish, Louisiana
Date of amendment request:
December 8, 2011, as supplemented by
letters dated April 11, May 2, and
September 5, 2012, and January 9 and
March 8, 2013.
Brief description of amendment: The
amendment revised Surveillance
Requirement (SR) 3.3.8.1.3 (calibration
of loss of power instrumentation) to
extend the frequency of the SR from 18
to 24 months, and revised certain
Allowable Values in TS 3.3.8.1, ‘‘Loss of
Power Instrumentation.’’
Date of issuance: March 29, 2013.
Effective date: As of the date of
issuance and shall be implemented 90
days from the date of issuance.
Amendment No.: 179.
Facility Operating License No. NPF–
47: The amendment revised the Facility
Operating License and Technical
Specifications.
Date of initial notice in Federal
Register: April 17, 2012 (77 FR 22811).
The supplemental letters dated April 11,
May 2, and September 5, 2012, and
January 9 and March 8, 2013, provided
additional information that clarified the
application, did not expand the scope of
the application as originally noticed,
and did not change the staff’s original
proposed no significant hazards
consideration determination as
published in the Federal Register.
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated March 29, 2013.
No significant hazards consideration
comments received: No.
Entergy Nuclear Operations, Inc.,
Docket No. 50–255, Palisades Nuclear
Plant, Van Buren County, Michigan
Date of application for amendment:
September 6, 2012.
Brief description of amendment: The
amendment revised the technical
specifications (TS) by adding a new
Limiting Condition for Operation (LCO)
3.0.8 associated with the impact of
inoperable snubbers. This LCO
establishes conditions under which TS
systems would remain operable when
required snubbers are not capable of
providing the related support function.
The proposed amendment is consistent
with NRC’s approved Technical
Specification Task Force (TSTF)
Improved Standard Technical
Specifications Change Traveler, TSTF–
372, Revision 4, ‘‘Addition of LCO 3.0.8,
Inoperability of Snubbers.’’
Date of issuance: March 29, 2013.
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Effective date: As of the date of
issuance and shall be implemented
within 90 days.
Amendment No.: 251.
Facility Operating License No. DPR–
20: Amendment revised the Technical
Specifications.
Date of initial notice in Federal
Register: November 27, 2012 (77 FR
70841).
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated March 29, 2013.
No significant hazards consideration
comments received: No.
Exelon Generation Company, LLC,
Docket Nos. STN 50–456 and STN 50–
457, Braidwood Station, Units 1 and 2,
Will County, Illinois
Docket Nos. STN 50–454 and STN 50–
455, Byron Station, Units 1 and 2, Ogle
County, Illinois
Date of application for amendment:
March 22, 2012, as supplemented by
letter dated December 3, 2012.
Brief description of amendment: The
proposed amendment would modify
technical specification (TS)
requirements regarding steam generator
tube inspections and reporting as
described in Technical Specifications
Task Force (TSTF)–510, ‘‘Revision to
Steam Generator Program Inspection
Frequencies and Tube Sample
Selection,’’ with proposed variations
and deviations.
Date of Issuance:. March 25, 2013.
Effective Date: As of the date of
issuance and shall be implemented
within 30 days.
Amendment Nos.: 172 and 170.
Facility Operating License Nos. NPF–
72, NPF–77, NPF–37, and NPF–66: The
amendments revised the TS and license.
Date of initial notice in Federal
Register: (77 FR 31660; May 29, 2012).
The December 3, 2012, supplement did
not increase the scope of the application
and did not change the NRC staff’s
initial proposed finding of no significant
hazards consideration.
The Commission’s related evaluation
of the amendments is contained in a
Safety Evaluation dated March 25, 2013.
No significant hazards consideration
comments received: No.
Exelon Generation Company, LLC, and
PSEG Nuclear LLC, Docket Nos. 50–277
and 50–278, Peach Bottom Atomic
Power Station, Units 2 and 3, York and
Lancaster Counties, Pennsylvania
Date of application for amendments:
April 27, 2012, as supplemented on
October 15, 2012.
Brief description of amendments: The
amendments: (1) Adopted a new
methodology for preparation of the
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reactor coolant system pressuretemperature (P–T) limits, (2) relocated
the P–T limits in the Technical
Specifications (TSs) to a new licenseecontrolled document, the Pressure and
Temperature Limits Report (PTLR), and
(3) modified the TSs to add references
to the PTLR.
Date of issuance: April 1, 2013.
Effective date: As of the date of
issuance, to be implemented within 60
days.
Amendments Nos.: 286 and 289.
Renewed Facility Operating License
Nos. DPR–44 and DPR–56: The
amendments revised the License and
TSs.
Date of initial notice in Federal
Register: July 3, 2012 (77 FR 39525).
The letter dated October 15, 2012,
provided clarifying information that did
not change the initial proposed no
significant hazards consideration
determination or expand the application
beyond the scope of the original Federal
Register notice.
The Commission’s related evaluation
of the amendments is contained in a
Safety Evaluation dated April 1, 2013.
No significant hazards consideration
comments received: No.
Florida Power and Light Company,
Docket Nos. 50–250 and 50–251, Turkey
Point Plant, Units 3 and 4, Miami-Dade
County, Florida
Date of application for amendments:
September 6, 2012, as supplemented by
letter dated January 11, 2013.
Brief description of amendments: The
amendments revised Technical
Specification (TS) 3/4.6.2.3,
‘‘Recirculation pH Control System and
NaTB Basket Minimum Loading
Requirement,’’ to reduce the minimum
loading requirement of sodium
tetraborate.
Date of issuance: April 2, 2013.
Effective date: As of the date of
issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 257 and 253.
Renewed Facility Operating License
Nos. DPR–31 and DPR–41: Amendments
revised the TSs.
Date of initial notice in Federal
Register: January 25, 2013 (78 FR 5505).
The supplement dated January 11, 2013,
provided additional information that
clarified the application, did not expand
the scope of the application as originally
noticed, and did not change the staff’s
original proposed no significant hazards
consideration determination as
published in the Federal Register.
The Commission’s related evaluation
of the amendments is contained in a
Safety Evaluation dated April 2, 2013.
No significant hazards consideration
comments received: No.
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Indiana Michigan Power Company,
Docket Nos. 50–315 and 50–316, Donald
C. Cook Nuclear Plant, Unit 1, Berrien
County, Michigan
Date of application for amendments:
September 12, 2012
Brief description of amendments: The
amendments revised the Technical
Specifications (TSs) to adopt NRCapproved TS Task Force (TSTF)
Traveler TSTF–510, Revision 2,
‘‘Revision to Steam Generator Program
Inspection Frequencies and Tube
Sample Selection,’’ using the
consolidated line item improvement
process. Specifically, the amendments
revise TS 3.4.17, ‘‘Steam Generator (SG)
Tube Integrity,’’ TS 5.5.7, ‘‘Steam
Generator (SG) Program,’’ and TS 5.6.7,
‘‘Steam Generator Tube Inspection
Report,’’ and include TS Bases changes
that summarize and clarify the purpose
of the TS.
Date of issuance: March 22, 2013.
Effective date: As of the date of
issuance and shall be implemented
within 180 days.
Amendment Nos.: 320 and 304.
Facility Operating License Nos. DPR–
58 and DPR–74: Amendments revise the
Operating Licenses and the Technical
Specifications.
Date of initial notice in Federal
Register: December 26, 2012 (77 FR
76080).
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated March 22, 2013.
No significant hazards consideration
comments received: No.
Omaha Public Power District, Docket
No. 50–285, Fort Calhoun Station, Unit
1, Washington County, Nebraska
Date of amendment request: March 9,
2012, as supplemented by letter dated
October 31, 2012.
Brief description of amendment: The
amendment relocated the Fort Calhoun
Station (FCS) Technical Specification
(TS) Limiting Condition of Operation
(LCO) 2.17, Miscellaneous Radioactive
Material Sources, and the associated
Surveillance Requirement (SR) 3.13,
Radioactive Material Sources
Surveillance, from the FCS TSs.
NUREG–1432, Revision 3, ‘‘Standard
Technical Specifications, Combustion
Engineering Plants,’’ does not contain a
TS or SR for radioactive source
surveillance. The operability and
surveillance requirements for leak
checking of miscellaneous radioactive
material sources will be incorporated
into the FCS Updated Safety Analysis
Report and associated plant procedures.
Date of issuance: March 29, 2013.
Effective date: As of its date of
issuance and shall be implemented
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22575
within 120 days from the date of
issuance.
Amendment No.: 271.
Renewed Facility Operating License
No. DPR–40: The amendment revised
the facility operating license and the
Technical Specifications.
Date of initial notice in Federal
Register: November 13, 2012 (77 FR
67684). The supplemental letter dated
October 31, 2012, provided additional
information that clarified the
application, did not expand the scope of
the application as originally noticed,
and did not change the staff’s original
proposed no significant hazards
consideration determination as
published in the Federal Register.
The Commission’s related evaluation
of the amendment is contained in a
safety evaluation dated March 29, 2013.
No significant hazards consideration
comments received: No.
PSEG Nuclear LLC, Docket No. 50–272,
Salem Nuclear Generating Station, Unit
1, Salem County, New Jersey
Date of application for amendment:
May 8, 2012.
Brief description of amendment: The
amendment revised Salem Unit 1
Technical Specification (TS) 6.8.4.i,
‘‘Steam Generator (SG) Program,’’ to
permanently exclude portions of the
tube below the top of the steam
generator tubesheet from periodic steam
generator tube inspections. In addition,
this amendment also revises TS
6.9.1.10, ‘‘Steam Generator Tube
Inspection Report,’’ to provide
permanent reporting requirements that
have been previously established on an
interim basis. The amendment was
submitted pursuant to 10 CFR 50.90,
‘‘Application for amendment of license,
construction permit, or early site
permit.’’
Date of issuance: March 28, 2013.
Effective date: The license
amendment is effective as of the date of
issuance and shall be implemented
within 60 days.
Amendment No.: 303.
Renewed Facility Operating License
No. DPR–70: The amendment revised
the facility operating license and the
Technical Specifications.
Date of initial notice in Federal
Register: January 22, 2013 (78 FR 4474).
The Commission’ related evaluation
of the amendments is contained in a
Safety Evaluation dated March 28, 2013.
No significant hazards consideration
comments received: No.
Dated at Rockville, Maryland, this 5th day
of April 2013.
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22576
Federal Register / Vol. 78, No. 73 / Tuesday, April 16, 2013 / Notices
For the Nuclear Regulatory Commission.
John D. Monninger,
Deputy Director, Division of Operating
Reactor Licensing, Office of Nuclear Reactor
Regulation.
[FR Doc. 2013–08756 Filed 4–15–13; 8:45 am]
BILLING CODE 7590–01–P
NUCLEAR REGULATORY
COMMISSION
[Docket No. 50–361; NRC–2013–0070]
Application and Amendment to Facility
Operating License Involving Proposed
No Significant Hazards Consideration
Determination; San Onofre Nuclear
Generating Station, Unit 2
Nuclear Regulatory
Commission.
ACTION: License amendment request;
opportunity to comment, request a
hearing, and petition for leave to
intervene.
AGENCY:
SUMMARY: The U.S. Nuclear Regulatory
Commission (NRC) is considering
issuance of an amendment to Facility
Operating License No. NPF–10, issued
to Southern California Edison (SCE, the
licensee), for operation of the San
Onofre Nuclear Generating Station
(SONGS), Unit 2. The proposed
amendment makes a temporary change
to the steam generator management
program and the license condition for
maximum power. For the duration of
Unit 2, Cycle 17, the proposed
amendment would change the terms
‘‘full range of normal operating
conditions’’ and ‘‘normal steady state
full power operation’’ and restricts
operation to 70 percent of the maximum
authorized power level. ‘‘Full range of
normal operating conditions’’ and
‘‘normal steady state full power
operation’’ shall be based upon the
steam generators being operated under
conditions associated with reactor core
power levels up to 70 percent Rated
Thermal Power (2406.6 megawatts
thermal).
Submit comments by May 16,
2013. Requests for a hearing or petition
for leave to intervene must be filed by
June 17, 2013.
ADDRESSES: You may submit comments
by any of the following methods (unless
this document describes a different
method for submitting comments on a
specific subject):
• Federal Rulemaking Web site: Go to
https://www.regulations.gov and search
for Docket ID NRC–2013–0070. Address
questions about NRC dockets to Carol
Gallagher; telephone: 301–492–3668;
email: Carol.Gallagher@nrc.gov. For
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DATES:
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technical questions, contact the
individual(s) listed in the FOR FURTHER
INFORMATION CONTACT section of this
document.
• Mail comments to: Cindy Bladey,
Chief, Rules, Announcements, and
Directives Branch (RADB), Office of
Administration, Mail Stop: TWB–05–
B01M, U.S. Nuclear Regulatory
Commission, Washington, DC 20555–
0001.
• Fax comments to: RADB at 301–
492–3446.
For additional direction on accessing
information and submitting comments,
see ‘‘Accessing Information and
Submitting Comments’’ in the
SUPPLEMENTARY INFORMATION section of
this document.
FOR FURTHER INFORMATION CONTACT:
Brian Benney, Office of Nuclear Reactor
Regulation, U.S. Nuclear Regulatory
Commission, Washington, DC 20555;
telephone: 301–415–2767; email:
Brian.Benney@nrc.gov.
SUPPLEMENTARY INFORMATION:
I. Accessing Information and
Submitting Comments
A. Accessing Information
Please refer to Docket ID NRC–2013–
0070 when contacting the NRC about
the availability of information regarding
this document. You may access
information related to this document,
which the NRC possesses and is
publicly available, by any of the
following methods:
• Federal Rulemaking Web site: Go to
https://www.regulations.gov and search
for Docket ID NRC–2013–0070.
• NRC’s Agencywide Documents
Access and Management System
(ADAMS): You may access publicly
available documents online in the NRC
Library at https://www.nrc.gov/readingrm/adams.html. To begin the search,
select ‘‘ADAMS Public Documents’’ and
then select ‘‘Begin Web-based ADAMS
Search.’’ For problems with ADAMS,
please contact the NRC’s Public
Document Room (PDR) reference staff at
1–800–397–4209, 301–415–4737, or by
email to pdr.resource@nrc.gov. The
ADAMS accession number for each
document referenced in this notice (if
that document is available in ADAMS)
is provided the first time that a
document is referenced. The application
for amendment, dated April 5, 2013, as
supplemented on April 9, 2013, is
available in ADAMS under Accession
Nos. ML13098A043 and ML13100A021,
respectively.
• NRC’s PDR: You may examine and
purchase copies of public documents at
the NRC’s PDR, Room O1–F21, One
PO 00000
Frm 00066
Fmt 4703
Sfmt 4703
White Flint North, 11555 Rockville
Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC–2013–
0070 in the subject line of your
comment submission, in order to ensure
that the NRC is able to make your
comment submission available to the
public in this docket.
The NRC cautions you not to include
identifying or contact information that
you do not want to be publicly
disclosed in you comment submission.
The NRC will post all comment
submissions at https://
www.regulations.gov as well as enter the
comment submissions into ADAMS.
The NRC does not routinely edit
comment submissions to remove
identifying or contact information.
If you are requesting or aggregating
comments from other persons for
submission to the NRC, then you should
inform those persons not to include
identifying or contact information that
they do not want to be publicly
disclosed in their comment submission.
Your request should state that the NRC
does not routinely edit comment
submissions to remove such information
before making the comment
submissions available to the public or
entering the comment submissions into
ADAMS.
II. Introduction
The NRC is considering issuance of an
amendment to Facility Operating
License No. NPF–10, issued to SCE, for
operation of SONGS Unit 2, located in
San Diego County, California.
The proposed amendment makes a
temporary change to the steam generator
management program and the license
condition for maximum power. For the
duration of Unit 2, Cycle 17, the
proposed amendment would change the
terms ‘‘full range of normal operating
conditions’’ and ‘‘normal steady state
full power operation’’ and restricts
operation to 70 percent of the maximum
authorized power level. ‘‘Full range of
normal operating conditions’’ and
‘‘normal steady state full power
operation’’ shall be based upon the
steam generators being operated under
conditions associated with reactor core
power levels up to 70 percent Rated
Thermal Power (2406.6 megawatts
thermal). Before any issuance of the
proposed license amendment, the
Commission will have made findings
required by the Atomic Energy Act of
1954, as amended (the Act), and the
Commission’s regulations.
The Commission has made a
proposed determination that the license
amendment request involves no
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Agencies
[Federal Register Volume 78, Number 73 (Tuesday, April 16, 2013)]
[Notices]
[Pages 22563-22576]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2013-08756]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
[NRC-2013-0069]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
Background
Pursuant to Section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is
publishing this regular biweekly notice. The Act requires the
Commission publish notice of any amendments issued, or proposed to be
issued and grants the Commission the authority to issue and make
immediately effective any amendment to an operating license or combined
license, as applicable, upon a determination by the Commission that
such amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from March 21 to April 3, 2013. The last biweekly
notice was published on April 2, 2013 (78 FR 19746).
ADDRESSES: You may access information and comment submissions related
to this document, which the NRC possesses and is publicly-available, by
searching on https://www.regulations.gov under Docket ID NRC-2013-0069.
You may submit comments by any of the following methods:
Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket NRC-2013-0069. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: Carol.Gallagher@nrc.gov.
Mail comments to: Cindy Bladey, Chief, Rules,
Announcements, and Directives Branch (RADB), Office of Administration,
Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
Fax comments to: RADB at 301-492-3446.
For additional direction on accessing information and submitting
comments, see ``Accessing Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
SUPPLEMENTARY INFORMATION:
I. Accessing Information and Submitting Comments
A. Accessing Information
Please refer to Docket ID NRC-2013-0069 when contacting the NRC
about the availability of information regarding this document. You may
access information related to this document, which the NRC possesses
and is publicly available, by the following methods:
Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2013-0069.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may access publicly available documents online in the NRC
Library at https://www.nrc.gov/reading-rm/adams.html. To begin the
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to pdr.resource@nrc.gov. Documents may be viewed in
ADAMS by performing a search on the document date and docket number.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2013-0069 in the subject line of your
comment submission, in order to ensure that the NRC is able to make
your comment submission available to the public in this docket.
The NRC cautions you not to include identifying or contact
information in comment submissions that you do not want to be publicly
disclosed. The NRC posts all comment submissions at https://www.regulations.gov as well as entering the comment submissions into
ADAMS, and the NRC does not edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information in their comment submissions
that they do not want to be publicly disclosed. Your request should
state that the NRC will not edit comment submissions to remove such
information before making the comment submissions available to the
public or entering the comment submissions into ADAMS.
[[Page 22564]]
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses and Combined Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of Title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated; or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated; or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Within 60 days after the date of publication of this notice, any
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license or
combined license. Requests for a hearing and a petition for leave to
intervene shall be filed in accordance with the Commission's ``Agency
Rules of Practice and Procedure'' in 10 CFR Part 2. Interested
person(s) should consult a current copy of 10 CFR 2.309, which is
available at the NRC's PDR, located at One White Flint North, Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The
NRC regulations are accessible electronically from the NRC Library on
the NRC's Web site at https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is
filed by the above date, the Commission or a presiding officer
designated by the Commission or by the Chief Administrative Judge of
the Atomic Safety and Licensing Board Panel, will rule on the request
and/or petition; and the Secretary or the Chief Administrative Judge of
the Atomic Safety and Licensing Board will issue a notice of a hearing
or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the requestor/petitioner
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held. If
the final determination is that the amendment request involves no
significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment. If the final determination is that the amendment
request involves a significant hazards consideration, then any hearing
held would take place before the issuance of any amendment.
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC's E-Filing rule (72 FR 49139;
August 28, 2007). The E-Filing process requires participants to submit
and serve all adjudicatory documents over the internet, or in some
cases to mail copies on electronic storage media. Participants may not
submit paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at hearing.docket@nrc.gov, or by
telephone at 301-415-1677, to request (1) a digital identification (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign
[[Page 22565]]
documents and access the E-Submittal server for any proceeding in which
it is participating; and (2) advise the Secretary that the participant
will be submitting a request or petition for hearing (even in instances
in which the participant, or its counsel or representative, already
holds an NRC-issued digital ID certificate). Based upon this
information, the Secretary will establish an electronic docket for the
hearing in this proceeding if the Secretary has not already established
an electronic docket.
Information about applying for a digital ID certificate is
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing
the E-Submittal server are detailed in the NRC's ``Guidance for
Electronic Submission,'' which is available on the agency's public Web
site at https://www.nrc.gov/site-help/e-submittals.html. Participants
may attempt to use other software not listed on the Web site, but
should note that the NRC's E-Filing system does not support unlisted
software, and the NRC Meta System Help Desk will not be able to offer
assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC's
Web site. Further information on the Web-based submission form,
including the installation of the Web browser plug-in, is available on
the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with the NRC guidance
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the
documents are submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC's Office of the General Counsel and any
others who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC's Web site
at https://www.nrc.gov/site-help/e-submittals.html, by email at
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866 672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville,
Maryland, 20852, Attention: Rulemaking and Adjudications Staff.
Participants filing a document in this manner are responsible for
serving the document on all other participants. Filing is considered
complete by first-class mail as of the time of deposit in the mail, or
by courier, express mail, or expedited delivery service upon depositing
the document with the provider of the service. A presiding officer,
having granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
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Commission, or the presiding officer. Participants are requested not to
include personal privacy information, such as social security numbers,
home addresses, or home phone numbers in their filings, unless an NRC
regulation or other law requires submission of such information.
However, a request to intervene will require including information on
local residence in order to demonstrate a proximity assertion of
interest in the proceeding. With respect to copyrighted works, except
for limited excerpts that serve the purpose of the adjudicatory filings
and would constitute a Fair Use application, participants are requested
not to include copyrighted materials in their submission.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Requests for hearing,
petitions for leave to intervene, and motions for leave to file new or
amended contentions that are filed after the 60-day deadline will not
be entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the following three
factors in 10 CFR 2.309(c)(1): (i) The information upon which the
filing is based was not previously available; (ii) the information upon
which the filing is based is materially different from information
previously available; and (iii) the filing has been submitted in a
timely fashion based on the availability of the subsequent information.
For further details with respect to this license amendment
application, see the application for amendment, which is available for
public inspection at the NRC's PDR, located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. Publicly available documents created or received at the NRC are
accessible electronically through ADAMS in the NRC Library at https://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to
ADAMS or who encounter problems in accessing the documents located in
ADAMS should contact the NRC's PDR Reference staff at 1-800-397-4209,
301-415-4737, or by email to pdr.resource@nrc.gov.
Detroit Edison, Docket No. 50-341, Fermi 2, Monroe County, Michigan
Date of amendment request: January 11, 2013.
Description of amendment request: The proposed amendment would
revise Fermi 2 Technical Specifications (TS) to
[[Page 22566]]
incorporate the NRC-approved TSTF-423, Revision 1. The proposed
amendment would modify TS to risk-inform requirements regarding
selected Required Action end states by incorporating the boiling water
reactor (BWR) owner's group (BWROG) approved Topical Report NEDC-32988-
A, Revision 2, ``Technical Justification to Support Risk-Informed
Modification to Selected Required Action End States for BWR Plants.''
Additionally, the proposed amendment would modify the TS Required
Actions with a Note prohibiting the use of limiting condition for
operation (LCO) 3.0.4.a when entering the preferred end state (Mode 3)
on startup.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change allows a change to certain required end
states when the TS Completion Times for remaining in power operation
will be exceeded. Most of the requested technical specification (TS)
changes are to permit an end state of hot shutdown (Mode 3) rather
than an end state of cold shutdown (Mode 4) contained in the current
TS. The request was limited to: (1) Those end states where entry
into the shutdown mode is for a short interval, (2) entry is
initiated by inoperability of a single train of equipment or a
restriction on a plant operational parameter, unless otherwise
stated in the applicable TS, and (3) the primary purpose is to
correct the initiating condition and return to power operation as
soon as is practical. Risk insights from both the qualitative and
quantitative risk assessments were used in specific TS assessments.
Such assessments are documented in Section 6 of topical report NEDC-
32988-A, Revision 2, ``Technical Justification to Support Risk
Informed Modification to Selected Required Action End States for BWR
Plants.'' They provide an integrated discussion of deterministic and
probabilistic issues, focusing on specific TSs, which are used to
support the proposed TS end state and associated restrictions. The
NRC staff finds that the risk insights support the conclusions of
the specific TS assessments. Therefore, the probability of an
accident previously evaluated is not significantly increased, if at
all. The consequences of an accident after adopting TSTF-423 are no
different than the consequences of an accident prior to adopting
TSTF-423. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
If risk is assessed and managed, allowing a change to certain
required end states when the TS Completion Times for remaining in
power operation are exceeded (i.e., entry into hot shutdown rather
than cold shutdown to repair equipment) will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change and the commitment by the licensee to adhere to the guidance
in TSTF-IG-05-02, ``Implementation Guidance for TSTF-423, Revision
1, `Technical Specifications End States, NEDC-32988-A,'' will
further minimize possible concerns.
Thus, based on the above, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change allows, for some systems, entry into hot
shutdown rather than cold shutdown to repair equipment, if risk is
assessed and managed. The BWROG's risk assessment approach is
comprehensive and follows NRC staff guidance as documented in
Regulatory Guides (RG) 1.174 and 1.177. In addition, the analyses
show that the criteria of the three-tiered approach for allowing TS
changes are met. The risk impact of the proposed TS changes was
assessed following the three-tiered approach recommended in RG
1.177. A risk assessment was performed to justify the proposed TS
changes. The net change to the margin of safety is insignificant.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bruce R. Masters, DTE Energy, General
Counsel--Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-1279.
NRC Branch Chief: Robert D. Carlson.
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3 (ONS1, ONS2, and ONS3),
Oconee County, South Carolina
Date of amendment request: October 30, 2012.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TSs) to allow operation of a
reverse osmosis system during normal plant operation to purify the
water in the borated water storage tanks and the spent fuel pools.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change requests NRC's approval of design features
and controls that will be used to ensure that periodic limited
operation of a Reverse Osmosis (RO) System during Unit operation
does not significantly impact the Borated Water Storage Tank (BWST)
or Spent Fuel Pool (SFP) function or other plant equipment. The
proposed change also requests NRC to approve proposed Technical
Specification (TS) requirements that will impose operating
restrictions and isolation requirements on the RO System. Duke
Energy evaluated the effect of potential failures, identified
precautionary measures that must be taken before and during RO
System operation, and identified specific required operator actions
to protect affected structures, systems, and components (SSCs)
important to safety.
The new high energy piping and non-seismic piping being
installed for the RO System is non-QA1 and is postulated to fail and
cause an Auxiliary Building flood. Duke Energy determined that
adequate time is available to isolate the flood source (BWST or SFP)
prior to affecting SSCs important to safety.
The existing Auxiliary Building Flood evaluation postulates a
single break in the non-seismic piping occurring in a seismic event.
The addition of the RO System will not increase the probability of a
seismic event. The existing postulated source of the pipe break in
the Auxiliary Building is due to the piping not being seismically
designed. The new RO System piping is considered a potential source
of a single pipe break for the same reason. The new non-seismic RO
System piping is of similar quality as the existing non-seismic
piping and is no more likely to fail than the existing piping. As
such, the addition of new non-seismic piping does not significantly
increase the probability of occurrence of an Auxiliary Building
flood due to a single pipe break. An Auxiliary Building flood due to
a non-seismic RO
[[Page 22567]]
System pipe break does not increase the consequences of the flood
since the new non-seismic pipe break is bounded by the Auxiliary
Building flood caused by existing non-seismic pipe breaks.
Procedural controls will ensure that the boron concentration
does not go below the TS limit as a result of water returned from
the RO System with lower boron concentration. Thus, no adverse
effects from decreased boron concentration will occur.
The RO System takes suction from the top of the SFP to protect
SFP inventory. Plant procedures will prohibit the use of the RO
System for the Units 1 & 2 SFP during the time period directly after
an outage that requires the Units 1 & 2 SFP level to be maintained
higher than the TS Limiting Condition for Operation (LCO) 3.7.11
level requirement. The higher level is required to support TS LCO
3.10.1 requirements for Standby Shutdown Facility (SSF) Reactor
Coolant (RC) Makeup System operability (due to the additional decay
heat from the recently offloaded spent fuel). Plant procedures will
also specify the siphon be broken during this time period so the SFP
water above the RO suction point cannot be siphoned off if the RO
piping breaks. The proposed change does not impact the fuel
assemblies, the movement of fuel, or the movement of fuel shipping
casks. The SFP boron concentration, level, and temperature limits
will not be outside of required parameters due to restrictions/
requirements on the system's operation. In addition, the proposed
new TS will require the siphon be broken during movement of
irradiated fuel assemblies in the SFP or movement of cask over the
SFP. Therefore, RO System operation cannot occur during these
activities, effectively eliminating a Fuel Handling Accidents (FHA)
from occurring while the RO System is in operation.
The BWST is used for mitigation of Steam Generator Tube Rupture
(SGTR), Main Steam Line Break (MSLB), and Loss of Coolant Accidents
(LOCAs). The SGTR and MSLB are bounded by the small break (SBLOCA)
analyses with respect to the performance requirements for the High
Pressure Injection (HPI) System. In the normal mode of Unit
operation, the BWST is not an accident initiator. The SFP is
evaluated to maintain acceptable criticality margin for all abnormal
and accident conditions including FHAs and cask drop accidents. Both
the BWST and SFP are specified by TS requirements to have minimum
levels/volumes and boron concentrations. The BWST also has TS
requirements for temperature. Prior to RO System operation,
procedures will require the minimum required initial boron
concentration and initial level/volume to be adjusted. Additionally,
they will require the RO System to be operated for a specified
maximum time period before readjusting volume and boron
concentration prior to another RO session. This ensures that the TS
specified boron concentration and level/volume limits for both the
SFP and the BWST are not exceeded during RO System operation. Thus,
the design functions of the BWST and the SFP will continue to be met
during RO System operation.
Since the BWST and SFP will still have TS boron concentration
and level/volume requirements and the RO System will be isolated
prior to increasing radiation levels preventing access to the
isolation valve, the mitigation of a LOCA or FHA does not result in
an increase in dose consequence. Since the design basis LOCA
analysis for Oconee assumes 5 gpm back-leakage from the Reactor
Building sump to the BWST, the Emergency Operating Procedure will
require the RO System to be isolated from the BWST prior to switch
over to the recirculation phase. The proposed TS will require the RO
system to be isolated (by breaking the siphon) from the SFPs during
fuel handling activities and will require the seismic boundary valve
between the BWST and RO System to be OPERABLE in MODES 1, 2, 3, and
4.
The additional controls imposed by the proposed Technical
Specifications (TSs) will provide additional assurance that
isolation valves and operating restrictions credited to eliminate
the need to analyze new release pathways introduced by the RO system
will be in place.
Therefore, installation and operation of the RO System during
Unit operation and the proposed TS imposing operating restrictions
do not significantly increase the probability or consequences of any
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The RO System adds non-seismic piping in the Auxiliary Building.
However, the break of a single non-seismic pipe in the Auxiliary
Building has already been postulated as an event in the licensing
basis. The RO System also does not create the possibility of a
seismic event concurrent with a LOCA since a seismic event is a
natural phenomena event. The RO System does not adversely affect the
Reactor Coolant System pressure boundary. The suction to the RO
System, when using the system for BWST purification, contains a
normally closed manual seismic boundary valve so the seismic design
criteria is met for separation of seismic/non-seismic piping
boundaries.
Duke Energy also evaluated potential releases of radioactive
liquid to the environment due to RO System piping failures. Design
features, controls imposed by the proposed TS, and procedural
controls will preclude release of radioactive materials outside the
Auxiliary Building by ensuring the RO System will be isolated when
required.
The SFP suction line is designed such that the SFP water level
will not go below TS required levels, thus the fuel assemblies will
have the TS required water level over them. Procedural controls will
restrict the use of the RO System and require breaking vacuum on the
Units 1 & 2 SFP suction line when the SSF conditions require the SFP
level be raised to support SSF RC Makeup System operability. Thus,
the SFP water level will not be reduced below required water levels
for these conditions. RO System operating restrictions will prevent
reducing the SFP boron concentration below TS limits.
Since the BWST and SFP will still have TS boron concentration
and level/volume requirements and the RO System will be isolated
prior to increasing radiation levels preventing access to the
isolation valve, the mitigation of a LOCA or FHA does not result in
an increase in dose consequence. Since the design basis LOCA
analysis for Oconee assumes 5 gpm back-leakage from the Reactor
Building sump to the BWST, the Emergency Operating Procedure will
require the RO System to be isolated from the BWST prior to switch
over to the recirculation phase. The proposed TS will require the RO
system to be isolated (by breaking the siphon) from the SFPs prior
to movement of irradiated fuel assemblies in the SFP or movement of
cask over the SFP and will require the seismic boundary valve
between the BWST and RO System to be OPERABLE in MODES 1, 2, 3, and
4.
The additional controls imposed by the proposed TSs will provide
additional assurance that isolation valves and operating
restrictions credited to eliminate the need to analyze new release
pathways introduced by the RO system will be in place.
Therefore, operation of the RO System during Unit operation will
not create the possibility of a new or different kind of accident
from any kind of accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The RO System adds non-seismic piping in the Auxiliary Building.
Duke Energy evaluated the impact of RO System operation on SSCs
important to safety and determined that the proposed TS controls and
procedural controls will ensure that TS limits for SFP and BWST
volume, temperature, and boron concentration will continue to be met
during RO operation. For the BWST, these controls will ensure the TS
minimum BWST boron concentration and level are available to mitigate
the consequences of a small break LOCA or a large break LOCA. For
the SFP, these controls ensure the assumptions of the fuel handling
and cask drop accident analyses are preserved. Additionally, the
failure of non-seismic RO System piping will not significantly
impact SSCs important to safety. Oconee's licensing basis does not
assume a design basis event occurs simultaneously with a seismic
event. The proposed change does not significantly impact the
condition or performance of SSCs relied upon for accident
mitigation. This change does not alter the existing TS allowable
values or analytical limits. The existing operating margin between
Unit conditions and actual Unit setpoints is not significantly
reduced due to these changes. The assumptions and results in any
safety analyses are not impacted. Therefore, operation of the RO
System during Unit operation does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
[[Page 22568]]
Attorney for licensee: Lara S. Nichols, Associate General Counsel,
Duke Energy Corporation, 526 South Church Street-EC07H, Charlotte, NC
28202-1802.
NRC Branch Chief: Robert J. Pascarelli.
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Date of amendment request: February 22, 2013.
Description of amendment request: The proposed amendments would
revise the Technical Specification curves for pressure and temperature
limits on the reactor coolant system, and limits on heatup and cooldown
rates.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment replaces the current Oconee Nuclear
Station (ONS) Units 1, 2, and 3 pressure/temperature (P-T) limit
curves applicable to 33 effective full power years (EFPY) in
Technical Specification (TS) 3.4.3 with new P-T limit curves
applicable to 54 EFPY. The proposed changes also revise the Reactor
Coolant System heatup and cooldown rates and allowable reactor
coolant pump combinations of TS Tables 3.4.3-1 and 3.4.3-2. The
pressure-temperature (P-T) limit curves in the TSs were
conservatively generated in accordance with fracture toughness
requirements of ASME Code Section XI, Appendix G, and the minimum
pressure and temperature requirements as listed in Table 1 of 10 CFR
Part 50, Appendix G. The proposed changes do not impact the
capability of the reactor coolant pressure boundary (i.e., no change
in operating pressure, materials, seismic loading, etc.).
Therefore, the proposed changes do not increase the potential
for the occurrence of a loss of coolant accident (LOCA). The changes
do not modify the reactor coolant system pressure boundary, nor make
any physical changes to the facility design, material, or
construction standards. The probability of any design basis accident
(DBA) is not affected by this change, nor are the consequences of
any DBA affected by this change. The proposed P-T limits, heatup and
cooldown rates and allowable operating reactor coolant pump
combinations are not considered to be an initiator or contributor to
any accident analysis addressed in the ONS Updated Final Safety
Analyses Report (UFSAR).
The proposed changes will not impact assumptions and conditions
previously used in the radiological consequence evaluations nor
affect the mitigation of these consequences due to an accident
described in the UFSAR. Also, the proposed changes will not impact a
plant system such that previously analyzed SSCs might be more likely
to fail. The initiating conditions and assumptions for accidents
described in the UFSAR remain as analyzed.
Therefore, the probability or consequences of an accident
previously evaluated is not significantly increased.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The requirements for P-T limit curves have been in place since
the beginning of plant operation. The revised curves are based on a
later edition to Section XI of the ASME Code that incorporates
current industry standards for P-T curves. The revised curves are
based on reactor vessel irradiation damage predictions using
Regulatory Guide 1.99 methodology. No new failure modes are
identified nor are any SSCs required to be operated outside the
design bases.
Therefore, the possibility of a new or different kind of
accident from any kind of accident previously evaluated is not
created.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed P-T curves continue to maintain the safety margins
of 10 CFR Part 50, Appendix G, by defining the limits of operation
which prevent non-ductile failure of the reactor pressure vessel.
Analyses have demonstrated that the fracture toughness requirements
are satisfied and that conservative operating restrictions are
maintained for the purpose of low temperature overpressure
protection. The P-T limit curves provide assurance that the RCS
pressure boundary will behave in a ductile manner and that the
probability of a rapidly propagating fracture is minimized.
Therefore, this request does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lara S. Nichols, Deputy General Counsel,
Duke Energy Corporation, 526 South Church Street-EC07H, Charlotte, NC
28202-1802.
NRC Branch Chief: Robert J. Pascarelli.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois
Docket Nos. STN 50-454 and STN 50-455, Byron Station, Units 1 and 2,
Ogle County, Illinois
Date of amendment request: December 21, 2012.
Description of amendment request: The proposed amendment would
Revise Technical Specifications (TS) 3.3.6, ``Containment Ventilation
Isolation Instrumentation.'' Specifically, this amendment request
proposes to revise Footnote (b) of TS Table 3.3.6-1, ``Containment
Ventilation Isolation Instrumentation,'' which specifies the
``Containment Radiation--High'' trip setpoint for two containment area
radiation monitors (i.e., 1(2) RE-AR011 and 1(2) RE-AR012). The
proposed changes would revise the ``Containment Radiation--High'' trip
setpoint from the current, overly conservative value (i.e., a
submersion dose rate of less than or equal to 10 mRhr in the
containment building), to less than or equal to 2 times the containment
building background radiation reading at rated thermal power, which is
consistent with NUREG-1431, ``Standard Technical Specifications,
Westinghouse Plants.'' Upon reaching the ``Containment Radiation--
High'' setpoint, these area radiation monitors provide an isolation
signal to the containment normal purge, mini purge and post-LOCA (Loss
of Coolant Accident) systems' containment isolation valves.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The containment ventilation isolation radiation monitors serve
two primary functions, they:
a. act as backup to the SI [safety injection] signal to ensure
closing of the purge valves; and
b. are the primary means for automatically isolating containment
in the event of a fuel handling accident in containment.
Upon sensing a high radiation condition in containment, these
area radiation monitors provide an isolation signal to the
containment normal purge, mini purge and post- LOCA systems
containment isolation valves (i.e., a containment ventilation
isolation signal).
The accidents that could potentially be impacted by the proposed
change were evaluated; specifically the Loss of Coolant Accident
(LOCA), Control Rod Ejection Accident (CREA) and Fuel Handling
Accident (FHA) in Containment. The proposed change has no impact on
probability of these accidents occurring as the subject containment
radiation area monitors detect a high radiation condition resulting
from these accidents. The radiation monitors do not initiate any
accidents or transients. Changing the ``Containment
[[Page 22569]]
Radiation--High'' trip setpoint from ``<=10 mR/hr in the containment
building,'' to ``<=2 times the containment building background
radiation reading at rated thermal power'' only affects the point
(i.e., the radiation level in containment) at which a containment
ventilation isolation signal would be generated. The requested
change does not involve any physical plant modifications or
operational changes that could adversely affect system reliability
or performance of the radiation monitors, or that could affect the
probability of operator error.
The requested change does not affect any postulated accident
precursors and therefore, will not affect the probability of an
accident previously evaluated.
The proposed change was evaluated to determine the impact on the
dose consequences of the LOCA, CREA, or FHA in containment. The
evaluation assumed a containment purge was in progress at the onset
of the subject accidents and showed that the proposed change in the
containment radiation monitors' setpoint had no effect on the purge
valve isolation time. With regard to the LOCA and CREA, the safety
analysis assumes a prompt purge valve isolation time (i.e.,
approximately 60 seconds) that significantly bounds the radiation
monitor sensing and response time, and actual valve design closure
time (i.e., a total of approximately 7 seconds). The radiation
monitor setpoint is not considered in the safety analysis and any
dose contribution associated with the containment purge, due to the
proposed change in setpoint, was shown to be unaffected; therefore,
the proposed change has no impact on the already insignificant dose
contribution attributed to a containment purge during an accident of
less than one mrem.
The dose consequences associated with the FHA in containment are
also not impacted by the proposed change in containment radiation
monitor setpoint. The existing dose consequences resulting from a
FHA with moving non-RECENTLY IRRADIATED FUEL (i.e., fuel moved more
than 48 hours after reactor shutdown) conservatively assume the
containment purge valves remain open throughout the event;
therefore, a change in the isolation setpoint does not impact the
results of this analysis. With regard to movement of RECENTLY
IRRADIATED FUEL (i.e., fuel moved less then 48 hours after reactor
shutdown), EGC's [Exelon Generation Company] proposal deletes TS LCO
[limiting condition for operation] 3.9.4.c.2 which allowed the
containment purge valves to be open provided the containment
radiation isolation system is OPERABLE. Deletion of TS LCO 3.9.4.c.2
ensures that the containment purge valves are in the closed position
when moving RECENTLY IRRADIATED FUEL, thus removing dependence on
the containment radiation isolation system and associated radiation
monitor setpoint from the FHA dose consequences.
The four other additional TS changes associated with the
deletion of LCO 3.9.4, Item c.2, proposed for consistency (i.e.,
deleting a NOTE regarding MODE applicability, deleting a CONDITION
related only to LCO 3.9.4.c.2, deleting a footnote regarding MODE
applicability; and deleting two surveillances related to LCO
3.9.4.c.2), also have no affect on either the probability or
consequences of an accident previously evaluated.
Based on the above discussion, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not result in a change to the design of
the Containment Ventilation Isolation System or the manner in which
the system operates or provides plant protection. The containment
radiation monitors will sense radiation levels in the same way and
will respond in the same manner when the setpoint is exceeded. The
change in the ``Containment Radiation--High'' setpoint does not
create a new failure mode for the associated containment radiation
monitors or for any other plant equipment. The deletion of LCO
3.9.4, Item c.2, in support of the setpoint change during refueling
operations, is more conservative than the current allowances and
actually eliminates a potential failure mode for the assumed open
containment ventilation isolation valves as the proposed deletion of
LCO 3.9.4, Item c.2 would require the valves to be closed prior to
moving RECENTLY IRRADIATED FUEL.
The changes do not result in the creation of any new accident
precursors, the creation of any changes to the existing accident
scenarios, nor do they create any new or different accident
scenarios. Subsequently, the accidents defined in the UFSAR [updated
final safety analysis report] continue to represent the credible
spectrum of events to be analyzed which determine safe plant
operation.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The analysis methodologies used in the subject safety analyses
are not modified as a result of the proposed TS changes to the
``Containment Radiation--High'' trip setpoint or the deletion of LCO
3.9.4, Item c.2, or any of the other four associated TS changes.
Although the ``Containment Radiation--High'' trip setpoint is being
increased, the increase in response time to a high radiation
condition in containment, when compared to the current setpoint, is
negligible due to the projected prompt rise in containment radiation
level upon initiation of a LOCA. The dose consequences and resultant
margin of safety to the regulatory acceptance limits, due to
revising the ``Containment Radiation--High'' setpoint to <= 2 times
the containment building background radiation reading at rated
thermal power, was shown to be unaffected for normal at-power
containment releases; have a negligible impact on the associated
LOCA and CREA accident dose consequences; and have no impact on the
FHA when moving RECENTLY IRRADIATED FUEL. Therefore, the proposed
changes do not impact any analysis margins.
The proposed changes do not alter the manner in which the safety
limits, limiting safety system setpoints, or limiting conditions for
operation are determined. The current safety analyses remain
bounding since their conclusions are not affected by the proposed
changes. The safety systems credited in the safety analyses will
continue to be available to perform their mitigation functions. All
protection signals credited as the primary or secondary accident
mitigating functions, and all operator actions credited in the
accident analyses remain the same. The proposed changes will not
result in plant operation in a configuration outside the design
basis.
Based on the above information, the proposed change does not
result in a significant reduction in the margin of safety.
Based on the above evaluation, EGC concludes that the proposed
amendments do not involve a significant hazards consideration under
the standards set forth in 10 CFR 50.92, paragraph (c), and,
accordingly, a finding of no significant hazards consideration is
justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Acting Branch Chief: Jeremy S. Bowen.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Nuclear Generating Units 3 and 4, Miami-Dade County, Florida
Date of amendment request: January 29, 2013.
Description of amendment request: The license amendment request
proposes to remove completed and satisfied license conditions and to
correct inadvertent errors and incorrect references.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendments do not change or modify the fuel, fuel
handling processes, fuel storage racks, number of fuel assemblies
[[Page 22570]]
that may be stored in the spent fuel pool (SFP), decay heat
generation rate, or the spent fuel pool cooling and cleanup system.
The proposed amendments only limit crediting of burnable absorbers
in the spent fuel pool to Integrated Fuel Burnable Absorber (IFBA)
rods that were specifically addressed in the currently approved
criticality analysis ([Westinghouse Commercial Atomic Power report]
WCAP-1 7094-P, Revision 3). The removal of the phrase ``or an
equivalent amount of another burnable absorber'' eliminates the
possibility of crediting a burnable absorber other than IFBA for
storage of spent fuel assemblies in the spent fuel pool without
prior NRC's approval. The deletion of the license condition
associated with the Boraflex Remedy is editorial as it is no longer
applicable. The proposed amendments do not affect the ability of the
BAST [boric acid storage tank] to perform its function or the
ability of the CREVS [control room emergency ventilation system] to
perform its function. These latter proposed TS [technical
specification] changes correct inadvertent errors and are consistent
with the stated intent of original license submittals or delete
license conditions that are no longer applicable or that have been
fully satisfied.
The proposed amendments do not cause any physical change to the
existing spent fuel storage configuration, fuel makeup, RCS [reactor
coolant system] pressure boundary, reactor containment, or plant
systems. The proposed amendments do not affect any precursors to any
accident previously evaluated or do not affect any known mitigation
equipment or strategies.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendments do not change or modify the fuel, fuel
handling processes, fuel racks, number of fuel assemblies that may
be stored in the pool, decay heat generation rate, or the spent fuel
pool cooling and cleanup system. The proposed amendments do not
result in any changes to spent fuel or to fuel storage
configurations. The removal of the phrase ``or an equivalent amount
of another burnable absorber'' eliminates the possibility of
crediting a burnable absorber other than IFBA for storage of spent
fuel assemblies in the spent fuel pool without prior NRC approval.
The proposed amendments do not affect the ability of the BAST to
perform its function or the ability of the CREVS to perform its
function. These latter proposed TS changes correct inadvertent
errors and are consistent with the stated intent of the original
license submittals, delete license conditions that are no longer
applicable or have been fully satisfied.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in the margin of safety?
Response: No.
The proposed amendments do not change or modify the fuel, fuel
handling processes, fuel racks, number of fuel assemblies that may
be stored in the pool, decay heat generation rate, or the spent fuel
pool cooling and cleanup system. Therefore, the proposed amendments
have no impact to the existing margin of safety for subcriticality
required by 10 CFR 50.68(b)(4). The other proposed OL [operating
license] & TS changes correct inadvertent errors and are consistent
with the stated intent of the original license submittals or delete
license conditions that are no longer applicable or have been fully
satisfied.
Therefore, the proposed amendments do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: James Petro, Managing Attorney--Nuclear,
Florida Power & Light, P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: Jessie F. Quichocho.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station,
Nemaha County, Nebraska
Date of amendment request: June 25, 2012.
Description of amendment request: The amendment would revise the
description of the Fuel Handling Accident (FHA) in Section XIV-6.4 of
the Cooper Nuclear Station (CNS) Updated Safety Analysis Report (USAR).
The revised USAR FHA description is based on changes to the Design
Basis Accident FHA dose calculation, to reflect a 24-month fuel cycle
source term using a Global Nuclear Fuels (GNF) 10 x 10 fuel array,
reduce the bounding Radial Peaking Factor, and revise the total
effective dose equivalent (TEDE) contribution to consider the shine
contribution from external sources.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The analyses changes described by this proposed change to the
USAR are not initiators to events, and, therefore, do not involve
the probability of an accident. The changes to the FHA calculation
for radiological dose following a FHA incorporate the following:
--accounts for the increase to the source term owing to the use of
Global Nuclear Fuels (GNF) 10 x 10 fuel exposed over a 24-month fuel
cycle,
--reduces the Radial Peaking Factor from 2.0 to 1.95, and
--uses a calculated Control Room shine contribution that is added to
the FHA dose consequences.
The NRC computer code RADTRAD Version 3.03 is used for the
offsite and Control Room dose calculation. The RADTRAD code was
approved for use with the CNS FHA alternative source term (AST) dose
calculation in License Amendment 222.
Because the analysis affected by the changes are not considered
to be an initiator to any previously analyzed accident, these
changes cannot increase the probability of any previously evaluated
accident. Therefore, these changes do not increase the probability
of occurrence of an accident evaluated previously in the USAR.
The changes in FHA dose consequences to the Control Room
occupant resulting from the 24-month cycle/GNF 10 x 10 source term
(without crediting the offset by a reduced Radial Peaking Factor),
results in more than a minimal increase in the consequences of an
accident previously evaluated in the USAR, as stated in 10 CFR
50.59(c)(2)(iii). However, the resultant dose remains well within
the regulatory limits of 10 CFR 50.67. When the reduced Radial
Peaking Factor is applied, the dose consequences are minor.
Therefore, this change does not significantly increase the
consequences of an accident previously evaluated in the USAR.
In summary, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This change does not involve initiators to any events in the
USAR, nor does the activity create the possibility for any new
accidents. Rather, this change is a result of the evaluation of the
most limiting FHA, which can occur at CNS. The changes to the FHA
calculation for radiological dose following a FHA incorporate the
following:
--accounts for the increase to the source term owing to the use of
GNF 10 x 10 fuel exposed over a 24-month fuel cycle,
--reduces the Radial Peaking Factor from 2.0 to 1.95, and
--uses a calculated Control Room shine contribution that is added to
the FHA dose consequences.
The RADTRAD code accommodates the use of GNF 10 x 10 fuel
exposed over a 24-month fuel cycle in calculating the FHA dose
consequences. The reduction in Radial Peaking Factor remains
bounding over the
[[Page 22571]]
CNS core design. The calculated Control Room shine contribution
replaces the previously approved qualitative assessment. The
proposed change does not introduce any new modes of plant operation
and does not involve physical modifications to the plant. As a
result, no new failure modes are being introduced.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The dose consequences are calculated in accordance with the
regulatory guidance found in RG 1.183. The RADTRAD code was used, as
approved for application at CNS with License Amendment 222. With the
reduced Radial Peaking Factor applied to the GNF 10 x 10 fuel that
has been exposed over a 24-month fuel cycle, the dose consequences
are minor. The calculated shine contribution being added to the
total Control Room occupant FHA dose results are less than the
previous qualitative assessment results that are being replaced.
Accordingly, the safety margins to the regulatory dose limits are
preserved.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Branch Chief: Michael T. Markley.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center, Linn County, Iowa
Date of amendment request: November 13, 2012.
Description of amendment request: The proposed amendment would
revise Renewed Facility Operating License (RFOL) Condition C.12 to
clarify that the programs and activities, identified in Appendix A of
NUREG-1955 and the Updated Final Safety Analysis Report (UFSAR)
supplement are to be completed before the period of extended operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The amendment does not significantly increase the probability of
an accident since it does not involve a change to any plant
equipment that initiates a plant accident. The change clarifies
RFOLC [RFOL Condition] C.12. The license conditions deal with
administrative controls over information contained in the Updated
Final Safety Analysis Repo[r]t (UFSAR) supplement. The proposed
changes are administrative and the license conditions are not an
initiator or mitigator of any previously evaluated accidents.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously evaluated
since it does not involve any physical alteration of plant equipment
and does not change the method by which any safety-related system
performs its function. The license conditions deal with
administrative controls over information contained in the UFSAR
supplement. No new or different types of equipment will be installed
and the basic operation of installed equipment is unchanged.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment does not affect design codes or design
margins. The change that clarifies RFOLC C.12 is administrative in
nature and does not have the ability to affect analyzed safety
margins.
Therefore, operation of DAEC in accordance with the proposed
amendment will not involve a significant reduction in the margin to
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. James Petro, P. O. Box 14000, Juno
Beach, FL 33408-0420.
NRC Branch Chief: Robert D. Carlson.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center, Linn County, Iowa
Date of amendment request: December 21, 2012.
Description of amendment request: The proposed amendment would
modify the current DAEC Technical Specifications (TS) requirement to
operate the Standby Gas Treatment System for 10 hours on a frequency
specified in the Surveillance Frequency Control Program in accordance
with TSTF-522, Revision 0, ``Revise Ventilation System Surveillance
Requirements to Operate for 10 hours per Month.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change replaces an existing Surveillance
Requirement to operate the SGT System equipped with electric heaters
for a continuous 10 hour period with a requirement to operate the
SGT System for 15 continuous minutes without the heaters operating.
In addition, the electrical heater output test in the VFTP
(Specification 5.5.7.e) is proposed to be deleted and a
corresponding change in the charcoal filter testing (Specification
5.5.7.c) be made to require the testing be conducted at a humidity
of at least 95% RH, which is more stringent than the current testing
requirement of 70% RH.
These systems are not accident initiators and therefore, these
changes do not involve a significant increase in the probability of
an accident. The proposed system and filter testing changes are
consistent with current regulatory guidance for these systems and
will continue to assure that these systems perform their design
function which may include mitigating accidents. Thus the change
does not involve a significant increase in the consequences of an
accident.
Therefore, it is concluded that this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change replaces an existing Surveillance
Requirement to operate the SGT System equipped with electric heaters
for a continuous 10 hour period with a requirement to operate the
systems for 15 continuous minutes without the heaters operating. In
addition, the electrical heater output test in the VFTP
(Specification 5.5.7.e) is proposed to be deleted and a
corresponding change in the charcoal filter testing (Specification
5.5.7.c) be made to require the testing be conducted at a humidity
of at least 95% RH, which is more stringent than the current testing
requirement of 70% RH.
The change proposed for this ventilation system does not change
any system
[[Page 22572]]
operations or maintenance activities. Testing requirements will be
revised and will continue to demonstrate that the Limiting
Conditions for Operation are met and the system components are
capable of performing their intended safety functions. The change
does not create new failure modes or mechanisms and no new accident
precursors are generated.
Therefore, it is concluded that this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change replaces an existing Surveillance
Requirement to operate the SGT System equipped with electric heaters
for a continuous 10 hour period with a requirement to operate the
systems for 15 continuous minutes without heaters operating. In
addition, the electrical heater output test in the VFTP is proposed
to be deleted and a corresponding change in the charcoal filter
testing be made to require the testing be conducted at a humidity of
at least 95% RH, which is more stringent than the current testing
requirement of 70% RH.
The proposed increase to 95% RH in the required testing of the
charcoal filters compensates for the function of the heaters, which
was to reduce the humidity of the incoming air to below the
currently-specified value of 70% RH for the charcoal. The proposed
change is consistent with regulatory guidance and continues to
ensure that the performance of the charcoal filters is acceptable.
Therefore, it is concluded that this change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. James Petro, P.O. Box 14000, Juno Beach,
FL 33408-0420.
NRC Branch Chief: Robert D. Carlson.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: April 20, 2012.
Description of amendment request: The proposed amendment would
revise the TS 3.1.7 to approve the use of an alternative method, other
than the current method of the use of movable incore detectors system,
to monitor the position of control rod or shutdown rod, in the event of
a malfunction of the microprocessor rod position indication (MRPI)
system. The use of this alternative method would reduce the required
frequency of flux mapping using the movable incore detector system to
determine the position of the control or shutdown rod position that is
not being indicated. This will reduce the wear on the movable incore
detector system that is also used to complete other required TS
surveillances.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change provides an alternative method for verifying
rod position of one rod. The proposed change meets the intent of the
current specification in that it ensures verification of position of
the rod once every 8 hours. The proposed change provides only an
alternative method of monitoring rod position and does not change
the assumptions or results of any previously evaluated accident.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change provides only an alternative method of
determining the position of one rod. No new accident initiators are
introduced by the proposed alternative manner of performing rod
position verification. The proposed change does not affect the
reactor protection system. Hence, no new failure modes are created
that would cause a new or different kind of accidents from any
accident previously evaluated.
Therefore, operation of the facility in accordance with the
proposed amendments would not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The basis of TS 3.1.7 states that the operability of the rod
position indicators is required to determine control rod positions
and thereby ensure compliance with the control rod alignment and
insertion limits. The proposed change does not alter the requirement
to determine rod position but provides an alternative method for
determining the position of the affected rod. As a result, the
initial conditions of the accident analysis are preserved and the
consequences of previously analyzed accidents are unaffected.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear
Generation, Constellation Group, LLC, 750 East Pratt Street, 17 Floor,
Baltimore, MD 21202.
NRC Acting Branch Chief: Sean Meighan.
South Carolina Electric and Gas Docket Nos.: 52-027 and 52-028, Virgil
C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield County,
South Carolina
Date of amendment request: March 26, 2013.
Description of amendment request: The proposed change would amend
Combined License Nos.: NPF-93 and NPF-94 for Virgil C. Summer Nuclear
Station (VCSNS) Units 2 and 3 by departing from the plant-specific
design control document Tier 2* material contained within the Updated
Safety Analysis Report (UFSAR) by revising the structural criteria code
for anchoring of reinforcement bar within the nuclear island walls.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The design functions of the nuclear island structures are to
provide support, protection, and separation for the seismic Category
I mechanical and electrical equipment located in the nuclear island.
The nuclear island structures are structurally designed to meet
seismic Category I requirements as defined in Regulatory Guide 1.29.
The change of the requirements for anchoring headed
reinforcement does not have an adverse impact on the response of the
nuclear island structures to safe shutdown earthquake ground motions
or loads due to anticipated transients or postulated accident
conditions. The change of the requirements for anchoring headed
reinforcement does not impact the support, design, or operation of
mechanical and fluid systems. There is no change to plant systems or
the response of systems to postulated accident conditions. There is
no change to the predicted radioactive releases due to postulated
accident conditions. The plant response to previously evaluated
accidents or external events is not adversely affected, nor
[[Page 22573]]
does the change described create any new accident precursors.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change is to provide the requirements for anchoring
nuclear island headed reinforcement. The proposed change does not
change the design of the nuclear island structures except to a
limited extent to redistribute the shear reinforcement in the walls
of the nuclear island. The proposed change does not impact the
support, design, or operation of mechanical or fluid systems. The
proposed change does not result in a new failure mechanism for the
nuclear island structures or new accident precursors. As a result,
the design functions of the nuclear island structures and the
seismic Category I mechanical and electrical equipment located in
the nuclear island are not adversely affected by the proposed
change.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
No safety analysis or design basis acceptance limit/criterion is
challenged or exceeded by the proposed change, thus, no margin of
safety is reduced. The limited application of alternative criteria
for headed reinforcement does not reduce the margin of safety.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
NRC Branch Chief: Lawrence Burkhart, Acting.
Southern Nuclear Operating Company Docket Nos.: 52-025 and 52-026,
Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Burke County,
Georgia
Date of amendment request: March 20, 2013.
Description of amendment request: The proposed change would amend
Combined Licenses Nos.: NPF-91 and NPF-92 for Vogtle Electric
Generating Plant (VEGP) Units 3 and 4 by departing from the plant-
specific design control document Tier 2* material contained within the
Updated Safety Analysis Report (UFSAR) by revising the structural
criteria code for anchoring of reinforcement bar within the nuclear
island walls.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The design functions of the nuclear island structures are to
provide support, protection, and separation for the seismic Category
I mechanical and electrical equipment located in the nuclear island.
The nuclear island structures are structurally designed to meet
seismic Category I requirements as defined in Regulatory Guide 1.29.
The change of the requirements for anchoring headed
reinforcement does not have an adverse impact on the response of the
nuclear island structures to safe shutdown earthquake ground motions
or loads due to anticipated transients or postulated accident
conditions. The change of the requirements for anchoring headed
reinforcement does not impact the support, design, or operation of
mechanical and fluid systems. There is no change to plant systems or
the response of systems to postulated accident conditions. There is
no change to the predicted radioactive releases due to postulated
accident conditions. The plant response to previously evaluated
accidents or external events is not adversely affected, nor does the
change described create any new accident precursors.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change is to provide the requirements for anchoring
nuclear island headed reinforcement. The proposed change does not
change the design of the nuclear island structures except to a
limited extent to redistribute the shear reinforcement in the walls
of the nuclear island. The proposed change does not impact the
support, design, or operation of mechanical or fluid systems. The
proposed change does not result in a new failure mechanism for the
nuclear island structures or new accident precursors. As a result,
the design functions of the nuclear island structures and the
seismic Category I mechanical and electrical equipment located in
the nuclear island are not adversely affected by the proposed
change.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
No safety analysis or design basis acceptance limit/criterion is
challenged or exceeded by the proposed change, thus, no margin of
safety is reduced. The limited application of alternative criteria
for headed reinforcement does not reduce the margin of safety.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. M. Stanford Blanton, Blach & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Acting Branch Chief: Lawrence Burkhart.
Notice of Issuance of Amendments to Facility Operating Licenses and
Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
[[Page 22574]]
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the NRC's Public Document Room (PDR), located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. Publicly available documents created or received at the NRC are
accessible electronically through the Agencywide Documents Access and
Management System (ADAMS) in the NRC Library at https://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by email to
pdr.resource@nrc.gov.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit 2, New London County, Connecticut
Date of amendment request: December 17, 2012, as supplemented by
January 31, 2013.
Description of amendment request: The amendment revised the
Millstone Power Station, Unit 2 (MPS2) Technical Specification (TS)
Surveillance Requirement 4.4.3.2 to remove the requirement to perform
the quarterly surveillance for a pressurizer power-operated relief
valve (PORV) block valve that is being maintained closed in accordance
with TS 3.4.3 Action a. The proposed change is consistent with the
requirements of the Standard Technical Specification--Combustion
Engineering Plants (NUREG-1432, Revision 4).
Date of issuance: March 26, 2013.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 314.
Renewed Facility Operating License No. DPR-65: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: January 22, 2012 (78 FR
4472). The supplemental letter dated January 31, 2013, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 26, 2013.
No significant hazards consideration comments received: No.
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish,
Louisiana
Date of amendment request: December 8, 2011, as supplemented by
letters dated April 11, May 2, and September 5, 2012, and January 9 and
March 8, 2013.
Brief description of amendment: The amendment revised Surveillance
Requirement (SR) 3.3.8.1.3 (calibration of loss of power
instrumentation) to extend the frequency of the SR from 18 to 24
months, and revised certain Allowable Values in TS 3.3.8.1, ``Loss of
Power Instrumentation.''
Date of issuance: March 29, 2013.
Effective date: As of the date of issuance and shall be implemented
90 days from the date of issuance.
Amendment No.: 179.
Facility Operating License No. NPF-47: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: April 17, 2012 (77 FR
22811). The supplemental letters dated April 11, May 2, and September
5, 2012, and January 9 and March 8, 2013, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 29, 2013.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant, Van Buren County, Michigan
Date of application for amendment: September 6, 2012.
Brief description of amendment: The amendment revised the technical
specifications (TS) by adding a new Limiting Condition for Operation
(LCO) 3.0.8 associated with the impact of inoperable snubbers. This LCO
establishes conditions under which TS systems would remain operable
when required snubbers are not capable of providing the related support
function. The proposed amendment is consistent with NRC's approved
Technical Specification Task Force (TSTF) Improved Standard Technical
Specifications Change Traveler, TSTF-372, Revision 4, ``Addition of LCO
3.0.8, Inoperability of Snubbers.''
Date of issuance: March 29, 2013.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 251.
Facility Operating License No. DPR-20: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 27, 2012 (77
FR 70841).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 29, 2013.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois
Docket Nos. STN 50-454 and STN 50-455, Byron Station, Units 1 and
2, Ogle County, Illinois
Date of application for amendment: March 22, 2012, as supplemented
by letter dated December 3, 2012.
Brief description of amendment: The proposed amendment would modify
technical specification (TS) requirements regarding steam generator
tube inspections and reporting as described in Technical Specifications
Task Force (TSTF)-510, ``Revision to Steam Generator Program Inspection
Frequencies and Tube Sample Selection,'' with proposed variations and
deviations.
Date of Issuance:. March 25, 2013.
Effective Date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 172 and 170.
Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66:
The amendments revised the TS and license.
Date of initial notice in Federal Register: (77 FR 31660; May 29,
2012). The December 3, 2012, supplement did not increase the scope of
the application and did not change the NRC staff's initial proposed
finding of no significant hazards consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 25, 2013.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Date of application for amendments: April 27, 2012, as supplemented
on October 15, 2012.
Brief description of amendments: The amendments: (1) Adopted a new
methodology for preparation of the
[[Page 22575]]
reactor coolant system pressure-temperature (P-T) limits, (2) relocated
the P-T limits in the Technical Specifications (TSs) to a new licensee-
controlled document, the Pressure and Temperature Limits Report (PTLR),
and (3) modified the TSs to add references to the PTLR.
Date of issuance: April 1, 2013.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendments Nos.: 286 and 289.
Renewed Facility Operating License Nos. DPR-44 and DPR-56: The
amendments revised the License and TSs.
Date of initial notice in Federal Register: July 3, 2012 (77 FR
39525). The letter dated October 15, 2012, provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination or expand the application beyond
the scope of the original Federal Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 1, 2013.
No significant hazards consideration comments received: No.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of application for amendments: September 6, 2012, as
supplemented by letter dated January 11, 2013.
Brief description of amendments: The amendments revised Technical
Specification (TS) 3/4.6.2.3, ``Recirculation pH Control System and
NaTB Basket Minimum Loading Requirement,'' to reduce the minimum
loading requirement of sodium tetraborate.
Date of issuance: April 2, 2013.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 257 and 253.
Renewed Facility Operating License Nos. DPR-31 and DPR-41:
Amendments revised the TSs.
Date of initial notice in Federal Register: January 25, 2013 (78 FR
5505). The supplement dated January 11, 2013, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 2, 2013.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Unit 1, Berrien County, Michigan
Date of application for amendments: September 12, 2012
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) to adopt NRC-approved TS Task Force
(TSTF) Traveler TSTF-510, Revision 2, ``Revision to Steam Generator
Program Inspection Frequencies and Tube Sample Selection,'' using the
consolidated line item improvement process. Specifically, the
amendments revise TS 3.4.17, ``Steam Generator (SG) Tube Integrity,''
TS 5.5.7, ``Steam Generator (SG) Program,'' and TS 5.6.7, ``Steam
Generator Tube Inspection Report,'' and include TS Bases changes that
summarize and clarify the purpose of the TS.
Date of issuance: March 22, 2013.
Effective date: As of the date of issuance and shall be implemented
within 180 days.
Amendment Nos.: 320 and 304.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revise the Operating Licenses and the Technical Specifications.
Date of initial notice in Federal Register: December 26, 2012 (77
FR 76080).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 22, 2013.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit 1, Washington County, Nebraska
Date of amendment request: March 9, 2012, as supplemented by letter
dated October 31, 2012.
Brief description of amendment: The amendment relocated the Fort
Calhoun Station (FCS) Technical Specification (TS) Limiting Condition
of Operation (LCO) 2.17, Miscellaneous Radioactive Material Sources,
and the associated Surveillance Requirement (SR) 3.13, Radioactive
Material Sources Surveillance, from the FCS TSs. NUREG-1432, Revision
3, ``Standard Technical Specifications, Combustion Engineering
Plants,'' does not contain a TS or SR for radioactive source
surveillance. The operability and surveillance requirements for leak
checking of miscellaneous radioactive material sources will be
incorporated into the FCS Updated Safety Analysis Report and associated
plant procedures.
Date of issuance: March 29, 2013.
Effective date: As of its date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment No.: 271.
Renewed Facility Operating License No. DPR-40: The amendment
revised the facility operating license and the Technical
Specifications.
Date of initial notice in Federal Register: November 13, 2012 (77
FR 67684). The supplemental letter dated October 31, 2012, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated March 29, 2013.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-272, Salem Nuclear Generating Station,
Unit 1, Salem County, New Jersey
Date of application for amendment: May 8, 2012.
Brief description of amendment: The amendment revised Salem Unit 1
Technical Specification (TS) 6.8.4.i, ``Steam Generator (SG) Program,''
to permanently exclude portions of the tube below the top of the steam
generator tubesheet from periodic steam generator tube inspections. In
addition, this amendment also revises TS 6.9.1.10, ``Steam Generator
Tube Inspection Report,'' to provide permanent reporting requirements
that have been previously established on an interim basis. The
amendment was submitted pursuant to 10 CFR 50.90, ``Application for
amendment of license, construction permit, or early site permit.''
Date of issuance: March 28, 2013.
Effective date: The license amendment is effective as of the date
of issuance and shall be implemented within 60 days.
Amendment No.: 303.
Renewed Facility Operating License No. DPR-70: The amendment
revised the facility operating license and the Technical
Specifications.
Date of initial notice in Federal Register: January 22, 2013 (78 FR
4474).
The Commission' related evaluation of the amendments is contained
in a Safety Evaluation dated March 28, 2013.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 5th day of April 2013.
[[Page 22576]]
For the Nuclear Regulatory Commission.
John D. Monninger,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2013-08756 Filed 4-15-13; 8:45 am]
BILLING CODE 7590-01-P